WorldWideScience

Sample records for casks

  1. Cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system.

  2. Sodium-cooled LMFBR cask recommendations

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    In April of 1970 a design study to establish the parameters of a shipping cask for LMFBR Spent Fuel Assemblies from FFTF and the first demonstration plant was initiated. The basic criteria presented were that the cask should be limited to 75 tons, and that the cask should be compatible with the FFTF Fuel Assembly design and the first demonstration LMFBR Fuel Assembly design. Several features of the I-(182)-1 cask and their basis are described.

  3. COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.; CHASTAIN, S.A.

    2007-06-21

    The Disposable Solid Waste Cask (DSWC) is a shielded cask capable of transporting, storing, and disposing of six non-fuel core components or approximately 27 cubic feet of radioactive solid waste. Five existing DSWCs are candidates for use in storing and disposing of non-fuel core components and radioactive solid waste from the Interim Examination and Maintenance Cell, ultimately shipping them to the 200 West Area disposal site for burial. A series of inspections, studies, analyses, and modifications were performed to ensure that these casks can be used to safely ship solid waste. These inspections, studies, analyses, and modifications are summarized and attached in this report. Visual inspection of the casks interiors provided information with respect to condition of the casks inner liners. Because water was allowed to enter the casks for varying lengths of time, condition of the cask liner pipe to bottom plate weld was of concern. Based on the visual inspection and a corrosion study, it was concluded that four of the five casks can be used from a corrosion standpoint. Only DSWC S/N-004 would need additional inspection and analysis to determine its usefulness. The five remaining DSWCs underwent some modification to prepare them for use. The existing cask lifting inserts were found to be corroded and deemed unusable. New lifting anchor bolts were installed to replace the existing anchors. Alternate lift lugs were fabricated for use with the new lifting anchor bolts. The cask tiedown frame was modified to facilitate adjustment of the cask tiedowns. As a result of the above mentioned inspections, studies, analysis, and modifications, four of the five existing casks can be used to store and transport waste from the Interim Examination and Maintenance Cell to the disposal site for burial. The fifth cask, DSWC S/N-004, would require further inspections before it could be used.

  4. Radioactive fuel cask railcar humping study

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, L.T. (comp.)

    1978-01-01

    The response of two radioactive shipping casks due to railroad humping shocks was calculated using a spring-mass model. The two railcars for these casks had different coupling mechanisms and different tiedown arrangements. Humping tests had been performed on one of the railcars (ATMX-600) and the resulting shock spectra was used to adjust the spring-mass model to get matching results. One car (designed for cask shipment) was equipped with Freightmaster E-15 end of car coupler and had about /sup 1///sub 8/ in. free travel of the cask skid relative to the car. The other car (ATMX-600), equipped with Miner RF-333 draft gear, was designed for nuclear weapon shipment and adapted to nuclear waste shipment by fastening the casks to the floor. Both car frames were built by the same manufacturer and are very similar. The response of the casks was put in shock spectra format and a parametric study was performed with various cask weights. Additional studies were done on the effects of fastening the loose cask, and using the Freightmaster end of car coupler on the ATMX car. Half-sine response spectra were overlaid to include the natural frequency of the cask tiedown. The resulting shock amplitude was plotted against the cask weight for each car. The results show a constant acceleration level for all the weights on the car with hydraulic end-of-car coupler which results from constant force at that impact velocity. The cask acceleration can be reduced by fastening it to the car, rather than allowing it to move freely through some small space. This study also shows that the cask response can be optimized on railcars without hydraulic draft gear by adjusting the tiedown stiffness to keep the tiedown frequency different than car frequencies.

  5. Test Plan for Cask Identification Detector

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-29

    This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

  6. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-20

    Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

  7. Design review report FFTF interim storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.L.

    1995-01-03

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location.

  8. Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

    Energy Technology Data Exchange (ETDEWEB)

    Bare, Walter Claude; Ebner, Matthias Anthony; Torgerson, Laurence Dale

    2001-08-01

    This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft für Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion’s (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999.

  9. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  10. Cask Processing Enclosure Specification/Operation - 12231

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Ronald [Transuranic Waste Processing Center, Lenoir City TN, 37771 (United States)

    2012-07-01

    Following an evaluation of throughput rates in the Hot Cell at the Transuranic Waste Processing Center and considering the variability in the waste with respect to actual dose rates a new approach to processing transuranic waste was necessary. Compounding the issue was the remote equipment poor reliability and high down-time. After considering all the factors, the evaluation resulted in the design and construction of a new waste processing area for handling the concrete casks that predominately contain contact handled transuranic (TRU) waste. The area is called the Cask Processing Enclosure and essentially the Cask Processing Enclosure mimics the projects current process techniques used for processing Contact Handled -TRU waste in the existing Box Breakdown Area and Glovebox. The Cask Processing Enclosure approach was developed based on a review of the RH processing throughput rates in the Hot Cell. As the process was reviewed consideration was given to the variability in the waste with respect to actual dose rates and the lack of equipment reliability and high wear in the Hot Cell. Based on that review, a new contact handled processing area for handling the concrete casks is being constructed and startup is expected shortly following WM2012. The Cask Processing Enclosure essentially mimics the projects current process techniques used for processing Contact Handled waste in the existing Box Breakdown Area and Glovebox and the design takes into consideration six years of operational experience. (authors)

  11. Multiple-Angle Muon Radiography of a Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Durham, J. Matthew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Guardincerri, Elena [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Morris, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poulson, Daniel Cris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bacon, Jeffrey Darnell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Morley, Deborah Jean [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Plaud-Ramos, Kenie Omar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-23

    A partially loaded dry storage cask was imaged using cosmic ray muons. Since the cask is large relative to the size of the muon tracking detectors, the instruments were placed at nine different positions around the cask to record data covering the entire fuel basket. We show that this technique can detect the removal of a single fuel assembly from the center of the cask.

  12. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-06-01

    The Global Threat Reduction Initiative (GTRI) has many experiments yet to be irradiated in support of the High Performance Research Reactor fuels development program. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for post irradiation examination. To date, the General Electric (GE)-2000 cask has been used to transport GTRI experiments between these facilities. However, the availability of the GE-2000 cask to support future GTRI experiments is at risk. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger GTRI experiments. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled experiments. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask.

  13. Safety evaluation for packaging (onsite) SERF cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-10-24

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

  14. The Feasibility of Cask "Fingerprinting" as a Spent-Fuel, Dry-Storage Cask Safeguards Technique

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, K P; Vanier, P; Forman, L; Caffrey, G; Wharton, J; Lebrun, A

    2005-07-27

    This report documents a week-long measurement campaign conducted on six, dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. A gamma-ray imager, a thermal-neutron imager and a germanium spectrometer were used to collect data on the casks. The campaign was conducted to examine the feasibility of using the cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the germanium spectrometer and differences between thermal neutron images and neutron-dose meters, this result is thought to be due to the limitations of the extant imagers used, rather than of the basic concept. Results indicate that measurements with improved imagers could contain significantly more information. Follow-on measurements with new imagers either currently available as laboratory prototypes or under development are recommended.

  15. Inspection of Used Fuel Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Dennis C. Kunerth; Tim McJunkin; Mark McKay; Sasan Bakhtiari

    2012-09-01

    ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) regulates the storage of used nuclear fuel, which is now and will be increasingly placed in dry storage systems. Since a final disposition pathway is not defined, the fuel is expected to be maintained in dry storage well beyond the time frame originally intended. Due to knowledge gaps regarding the viability of current dry storage systems for long term use, efforts are underway to acquire the technical knowledge and tools required to understand the issues and verify the integrity of the dry storage system components. This report summarizes the initial efforts performed by researchers at Idaho National Laboratory and Argonne National Laboratory to identify and evaluate approaches to in-situ inspection dry storage casks. This task is complicated by the design of the current storage systems that severely restrict access to the casks.

  16. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    PRAGA, A.N.

    1998-10-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description.

  17. Safety analysis report for packaging: the ORNL loop transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Nelms, H.A.; Crowley, W.K.; Just, R.A.

    1977-11-01

    An evaluation of the ORNL loop transport cask demonstrating its compliance with the regulations governing the transportation of radioactive and fissile materials is presented. A previous review of the cask is updated to demonstrate compliance with current regulations, to present current procedures, and to reflect the more recent technology.

  18. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    Energy Technology Data Exchange (ETDEWEB)

    Duckett, A.J.; Sorenson, K.B.

    1985-09-01

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

  19. What drives Greek consumer preferences for cask wine?

    DEFF Research Database (Denmark)

    Chrysochou, Polymeros; Corsi, A. M.; Krystallis Krontalis, Athanasios

    2012-01-01

    to structure, the Greek cask wine market was found to consist of four distinct segments that were labelled as connoisseurs, convenience seekers, experienced and risk averse. These segments showed differences in relation to their past experience and in the importance given to intrinsic (quality, taste, origin......Purpose – Cask wine (bag-in-box, soft pack) has not received considerable attention in wine marketing research, but interest among winemakers and consumers has been increasing steadily. However, little is known about what drives consumer preferences for cask wine and, furthermore, what the profile...... of the cask wine consumer is. This study aims at filling this gap. Design/methodology/approach – Based on a web-based survey, the best-worst scaling (BWS) method was applied to measure the importance of attributes that Greek consumers assign when choosing cask wine. Then, a latent class clustering analysis...

  20. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  1. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  2. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each casks neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  3. Viability of Existing INL Facilities for Dry Storage Cask Handling

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randy; Wallace, Bruce; Winston, Phil; Marschman, Steve

    2013-04-30

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  4. Viability of Existing INL Facilities for Dry Storage Cask Handling

    Energy Technology Data Exchange (ETDEWEB)

    Randy Bohachek; Charles Park; Bruce Wallace; Phil Winston; Steve Marschman

    2013-04-01

    This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

  5. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. (Audin (Lindsay), Ossining, NY (United States))

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  6. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. [Audin (Lindsay), Ossining, NY (United States)

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  7. Nondestructive evaluation of LWR spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study.

  8. Feasibility study for a transportation operations system cask maintenance facility

    Energy Technology Data Exchange (ETDEWEB)

    Rennich, M.J.; Medley, L.G.; Attaway, C.R.

    1991-01-01

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the development of a waste management program for the disposition of spent nuclear fuel (SNF) and high-level waste (HLW). The program will include a transportation system for moving the nuclear waste from the sources to a geologic repository for permanent disposal. Specially designed casks will be used to safely transport the waste. The cask systems must be operated within limits imposed by DOE, the Nuclear Regulatory Commission (NRC), and the Department of Transportation (DOT). A dedicated facility for inspecting, testing, and maintaining the cask systems was recommended by the General Accounting Office (in 1979) as the best means of assuring their operational effectiveness and safety, as well as regulatory compliance. In November of 1987, OCRWM requested a feasibility study be made of a Cask Maintenance Facility (CMF) that would perform the required functions. 46 refs., 16 figs., 13 tabs.

  9. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  10. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (USA))

    1990-01-01

    The original cask designs for a cask demonstration project featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, however, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design. 6 refs., 2 figs., 2 tabs.

  11. Experience with certifying borated stainless steel as a shipping cask basket material

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States)

    1990-10-01

    This paper discusses the original cask designs for a cask demonstration project that has featured fuel baskets constructed of borated stainless steel (bss) as a structural material. The project is intended to demonstrate casks that can be used for both shipping and storing spent nuclear fuel assemblies. The baskets were intended to maintain the fuel assemblies in a subcritical array for both normal and accident conditions. The Nuclear Regulatory Commission, judged bss to be unacceptable as a structural material. The cask designs were subsequently modified. The knowledge gained during this cask demonstration project may be applicable to development of bss as a basket material in future cask design.

  12. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  13. TRANSPORTATION CASK RECEIPT AND RETURN FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    V. Arakali

    2005-02-24

    The purpose of this design calculation is to estimate radiation doses received by personnel working in the Transportation Cask Receipt and Return Facility (TCRRF) of the repository including the personnel at the security gate and cask staging areas. This calculation is required to support the preclosure safety analysis (PCSA) to ensure that the predicted doses are within the regulatory limits prescribed by the U.S. Nuclear Regulatory Commission (NRC). The Cask Receipt and Return Facility receives NRC licensed transportation casks loaded with spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TCRRF operation starts with the receipt, inspection, and survey of the casks at the security gate and the staging areas, and proceeds to the process facilities. The transportation casks arrive at the site via rail cars or trucks under the guidance of the national transportation system. This calculation was developed by the Environmental and Nuclear Engineering organization and is intended solely for the use of Design and Engineering in work regarding facility design. Environmental and Nuclear Engineering personnel should be consulted before using this calculation for purposes other than those stated herein or for use by individuals other than authorized personnel in the Environmental and Nuclear Engineering organization.

  14. Research on localization and alignment technology for transfer cask

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jingchuan, E-mail: jchwang@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China); Yang, Ming; Chen, Weidong [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China)

    2015-10-15

    Highlights: • A method for the alignment between TB and HCB based on localizability is proposed. • A localization method based on the localizability estimation is proposed to realize the cask's localization accurately and ensures the transfer cask's accurate docking in the front of the window of Tokmak Building. • The experimental results show that the proposed algorithm works well in the indoor simulation environment. This system will be test in EAST of China. - Abstract: According to the long length characteristics of transfer cask compared to the environment space between Tokmak Building (TB) and HCB (Hot Cell Building), this paper proposes an autonomous localization and alignment method for the internal components transportation and replacement. A localization method based on the localizability estimation is used to realize the cask's localization and navigation accurately. Once the cask arrives at the front of the TB window, the position and attitude measurement system is used to detect the relative alignment error between the seal door of pallet and the window of TB real-time. The alignment between seal door and TB window could be realized based on this offset. The simulation experiment based on the real model is designed according to the real TB situation. The experiment results show that the proposed localization and alignment method can be used for transfer cask.

  15. SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS

    Directory of Open Access Journals (Sweden)

    JAE-HUN KO

    2014-08-01

    The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2×10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2×10 cask array, dose rates at the center point of the array and at the center of the casks’ height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2×10 cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

  16. 77 FR 64834 - Computational Fluid Dynamics Best Practice Guidelines for Dry Cask Applications

    Science.gov (United States)

    2012-10-23

    ...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is requesting public comments on draft NUREG-2152, ``Computational Fluid Dynamics Best Practice Guidelines for Dry Cask Applications.'' The draft NUREG-2152 report provides best practice guidelines for undertaking simulations used to evaluate the thermal response of dry casks. Dry cask applications include transfer, transport, and......

  17. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... the Holtec International HI-STORM 100 dry cask storage system listing within the ``List of Approved... other aspects of the HI-STORM 100 dry storage cask system. Because the NRC considers this...

  18. Safety Assessment of a Metal Cask under Aircraft Engine Crash

    Directory of Open Access Journals (Sweden)

    Sanghoon Lee

    2016-04-01

    Full Text Available The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is freestanding on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact load–time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

  19. Safety assessment of a metal cask under aircraft engine crash

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Choi, Woo Seok; Seo, Ki Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is free standing on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact load-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

  20. 76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-06-08

    ... 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear... the NRC's spent fuel storage regulations to add the Holtec HI-STORM Flood/Wind cask system to the... Holtec HI- STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks''...

  1. Vestibule and Cask Preparation Mechanical Handling Calculation

    Energy Technology Data Exchange (ETDEWEB)

    N. Ambre

    2004-05-26

    The scope of this document is to develop the size, operational envelopes, and major requirements of the equipment to be used in the vestibule, cask preparation area, and the crane maintenance area of the Fuel Handling Facility. This calculation is intended to support the License Application (LA) submittal of December 2004, in accordance with the directive given by DOE correspondence received on the 27th of January 2004 entitled: ''Authorization for Bechtel SAIC Company L.L.C. to Include a Bare Fuel Handling Facility and Increased Aging Capacity in the License Application, Contract Number DE-AC28-01R W12101'' (Ref. 167124). This correspondence was appended by further correspondence received on the 19th of February 2004 entitled: ''Technical Direction to Bechtel SAIC Company L.L. C. for Surface Facility Improvements, Contract Number DE-AC28-01R W12101; TDL No. 04-024'' (Ref. 16875 1). These documents give the authorization for a Fuel Handling Facility to be included in the baseline. The limitations of this preliminary calculation lie within the assumptions of section 5 , as this calculation is part of an evolutionary design process.

  2. Standard review plan for dry cask storage systems. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements. Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 {open_quotes}Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask{close_quotes} contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.

  3. Contract Report for Usage Inspection of KN-12 Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

  4. ACCIDENTAL DROP OF A CARBON STEEL/LEAD SHIPPING CASK AT LOW TEMPERATURES

    Energy Technology Data Exchange (ETDEWEB)

    B. D. Hawkes; K. R. Durstine

    2007-07-01

    A shielded cask is used to transport radioactive materials between facilities. The cask was fabricated with an outer and inner shell of hot rolled low carbon steel. Lead was poured in the annular space between the shells to provide radiation shielding. Carbon steel is known to be susceptible to lowtemperature brittle fracture under impact loading. This paper will present the analysis results representing postulated transportation accidents during on-site transfers of the cask. The accident scenarios were based on a series of cask drops onto a rigid surface from a height of 6 ft assuming brittle failure of the cask shell at subzero temperatures. Finite element models of the cask and its contents were solved and post processed using ABAQUS software. Each model was examined for failure to contain radioactive materials and/or significant loss of radiation shielding. Results of these analyses show that the body of the cask exhibits considerable ruggedness and will remain largely intact after the impact. There will be deformation of the main cask body with localized brittle failure of the cask outer shell and components and but no complete penetration of the cask shielding. The cask payload outer waste can will experience some permanent plastic deformation in each drop, but will not be deformed to the point where it will rupture, thus maintaining confinement of the can contents.

  5. Benchmarking Data for the Proposed Signature of Used Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-23

    A set of benchmarking measurements to test facets of the proposed extended storage signature was conducted on May 17, 2016. The measurements were designed to test the overall concept of how the proposed signature can be used to identify a used fuel cask based only on the distribution of neutron sources within the cask. To simulate the distribution, 4 Cf-252 sources were chosen and arranged on a 3x3 grid in 3 different patterns and raw neutron totals counts were taken at 6 locations around the grid. This is a very simplified test of the typical geometry studied previously in simulation with simulated used nuclear fuel.

  6. Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

    Science.gov (United States)

    Carstens, Thomas Alan

    This study focuses on the development of self-powered wireless sensors. These sensors can be used to measure key parameters in extreme environments; e.g., temperature monitoring for spent nuclear fuel during dry-cask storage. This study has developed a design methodology for these self-powered monitoring systems. The main elements that constitute this work consist of selecting and testing a power source for the wireless sensor, determination of the attenuation of the wireless signal, and testing the wireless sensor circuitry in an extreme environment. OrigenArp determined the decay heat and gamma/neutron source strength of the spent fuel throughout the service life of the dry-cask. A first principles analysis modeled the temperatures inside the dry-cask. A finite-element heat transfer code calculated the temperature distribution of the thermoelectric and heat sink. The temperature distributions determine the power produced by the thermoelectric. It was experimentally verified that a thermoelectric generator (HZ-14) with a DC/DC converter (Linear Technology LTC3108EDE) can power a transceiver (EmbedRF) at condition which represent prototypical conditions throughout and beyond the service life of the dry-cask. The wireless sensor is required to broadcast with enough power to overcome the attenuation from the dry-cask. It will be important to minimize the attenuation of the signal in order to broadcast with a small transmission power. To investigate the signal transmission through the dry-cask, CST Microwave Studio was used to determine the scattering parameter S2,1 for a horizontal dry-cask. Important parameters that can influence the transmission of the signal are antenna orientation, antenna placement, and transmission frequency. The thermoelectric generator, DC/DC converter, and transceiver were exposed to 60Co gamma radiation (exposure rate170.3 Rad/min) at the University of Wisconsin Medical Radiation Research Center. The effects of gamma radiation on the

  7. Safety analysis report for medical radioisotope transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Ku, J. H.; Lee, J. C. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    KAERI has been producing radioisotopes for medical and industrial use and supplying them to radioisotope-using hospitals and industries. RI transport cask of A type package has been developed to transport medical radioisotopes from the HANARO to the hospitals. The safety analyses were performed under normal transport conditions in accordance with standards of transport regulations. As a results, it should be verified that the cask maintains the shielding and structural integrities under prescribed condition by the regulations. 8 refs., 20 figs., 7 tabs. (Author)

  8. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

  9. THERMAL MODELING ANALYSIS OF SRS 70 TON CASK

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.; Jordan, J.; Hensel, S.

    2011-03-08

    The primary objective of this work was to perform the thermal calculations to evaluate the Material Test Reactor (MTR) fuel assembly temperatures inside the SRS 70-Ton Cask loaded with various bundle powers. MTR fuel consists of HFBR, MURR, MIT, and NIST. The MURR fuel was used to develop a bounding case since it is the fuel with the highest heat load. The results will be provided for technical input for the SRS 70 Ton Cask Onsite Safety Assessment. The calculation results show that for the SRS 70 ton dry cask with 2750 watts total heat source with a maximum bundle heat of 670 watts and 9 bundles of MURR bounding fuel, the highest fuel assembly temperatures are below about 263 C. Maximum top surface temperature of the plastic cover is about 112 C, much lower than its melting temperature 260 C. For 12 bundles of MURR bounding fuel with 2750 watts total heat and a maximum fuel bundle of 482 watts, the highest fuel assembly temperatures are bounded by the 9 bundle case. The component temperatures of the cask were calculated by a three-dimensional computational fluid dynamics approach. The modeling calculations were performed by considering daily-averaged solar heat flux.

  10. Seismic Performance of Dry Casks Storage for Long- Term Exposure

    Energy Technology Data Exchange (ETDEWEB)

    Ibarra, Luis [Univ. of Utah, Salt Lake City, UT (United States); Sanders, David [Univ. of Nevada, Reno, NV (United States); Yang, Haori [Oregon State Univ., Corvallis, OR (United States); Pantelides, Chris [Univ. of Utah, Salt Lake City, UT (United States)

    2016-12-30

    The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred to DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks

  11. Accidental Drop of a Carbon Steel/Lead Shipping Cask (HFEF 14) at Low Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Brian D. Hawkes; Michael E. Nitzel

    2007-08-01

    A shielded cask is used to transport radioactive materials between facilities at the Idaho National Laboratory. The cask was fabricated with an outer and inner shell of A36 carbon steel with lead poured in the annular space between the shells to provide radiation shielding. Carbon steel is known to be susceptible to low-temperature brittle fracture under impact loading. This paper will present the analysis results representing postulated transportation accidents during on-site transfers of the cask at subzero temperatures. The accident scenarios were based on a series of cask drops onto a rigid surface from a height of 1.83m (6 ft.) Finite element models of the cask and its contents were solved and post processed using the ABAQUS software. Each model was examined for failure to contain radioactive materials and/or significant loss of radiation shielding. Results of these analyses show that the body of the cask exhibits considerable ruggedness and will remain largely intact after the impact. There will be deformation of the main cask body with localized brittle failure of the cask outer shell and door structure. The cask payload outer waste can remains in the cask but will experience some permanent plastic deformation in each drop. It will not be deformed to the point where it will rupture, thus maintaining confinement of the can contents.

  12. Design of casks: incorporating operational feedback from maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Bimet, F.; Hartenstein, M. [COGEMA Logistics, Saint Quentin (France)

    2004-07-01

    Casks are designed to conform to regulations and to client specifications. Essential areas such as easy operation, low costs of maintenance, low operation and maintenance doses, limited waste, are not explicitly covered. Notwithstanding, COGEMA LOGISTICS uses all feedback available, so that casks are designed to be easy, safe and economical to operate and maintain. Maintenance is an activity where you do spot items that old-time designers could have made better, and things that users should not have done. Thanks to quality assurance, there are a number of data available, waiting to be collected and exploited; they have to be identified, located, retrieved, and analysed. Using information such as wear, damage, use of spare parts, access problems helps to make casks ever better. It leads to more efficient concepts, and to upgrades on existing designs; it also allows to integrate environmental considerations, inter alia in the choice of materials and in maintenance methods. It is necessary for the designer to interact with the users, the cask owners, the maintenance providers, in order to gather all relevant information and events. This is made easier when all these actors are ''under one roof'', or have very close ties. This paper presents COGEMA LOGISTICS methods for collecting and analysing all these experiences waiting to be used. It explains our process for analysing data, preparing yearly reports that are made available to our designers. It describes how each new design is subject to a maintainability study, using this feedback, so that the cask safety is always assured, that radiological doses are kept to a minimum, and that operating and maintenance costs will remain as low as possible.

  13. Phenotypic and molecular insights into CASK-related disorders in males

    OpenAIRE

    Moog, U.; T Bierhals; Brand, K.; Bautsch, J.; Biskup, S.; Brune, T.; J. Denecke; Die-Smulders, C E M de; Evers, C.; Hempel, M.; Henneke, M; Yntema, H.G.; Menten, B.; Pietz, J.; Pfundt, R.P.

    2015-01-01

    Background Heterozygous loss-of-function mutations in the X-linked CASK gene cause progressive microcephaly with pontine and cerebellar hypoplasia (MICPCH) and severe intellectual disability (ID) in females. Different CASK mutations have also been reported in males. The associated phenotypes range from nonsyndromic ID to Ohtahara syndrome with cerebellar hypoplasia. However, the phenotypic spectrum in males has not been systematically evaluated to date. Methods We identified a CASK alteration...

  14. Design analysis report for the TN-WHC cask and transportation system

    Energy Technology Data Exchange (ETDEWEB)

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13

    This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

  15. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Science.gov (United States)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  16. Survivability Tests on a Nuclear Waste Cask in Simulated Railroad Accident Fires.

    Science.gov (United States)

    1983-06-01

    Test Number 1 42 11. The Wind Direction as a Function of Time During the HNPF Cask Thermal Test Number 1...43 12. The Wind Speed as a Function of Time During the HNPF Cask Thermal Test Number 1 .......................................... 44 13. The Ambient...60 27. A View of the HNPF Cask Taken During Torch Thermal Test Number 2 62 28. The Wind Direction as a

  17. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  18. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations.

  19. STABILITY EVALUATION OF METAL CASK ATTACHED TO A TRANSFER PALLET DURING LONG-PERIOD SEISMIC MOTIONS

    Science.gov (United States)

    Kawaguchi, Shohei; Shirai, Koji; Kanazawa, Kenji

    Rocking behavior of unfixed body is affected by center of mass, material coefficient of restitution and so on. 2/5 scale metal cask model considering these parameter was used for seismic test to evaluate stability of grounding metal cask attached to a transfer pallet under the influence of long-period earthquake motion. The newest knowledge from seismic test indicates seismic motion with high velocity over 100 kine not always cause the raise of response velocity of metal cask because of energy consumption by cask sliding and impact deformation of concrete. And new estimation method (called "Window energy spectrum method") of earthquake response spectrum gives suitable evaluation of response energy.

  20. Safety evaluation for packaging (onsite) disposable solid waste cask

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, B.D., Westinghouse Hanford

    1996-12-20

    This safety evaluation for packaging (SEP) evaluates and documents the ability of the Disposable Solid Waste Cask (DSWC) to meet the packaging requirements of HNF-CM-2-14, Hazardous Material Packaging and Shipping, for the onsite transfer of special form, highway route controlled quantity, Type B fissile radioactive material. This SEP evaluates five shipments of DSWCs used for the transport and storage of Fast Flux Test Facility unirradiated fuel to the Plutonium Finishing Plant Protected Area.

  1. Performance of bolted closure joint elastomers under cask aging conditions

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Daugherty, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-23

    The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.

  2. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  3. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration.

  4. 78 FR 8050 - Spent Fuel Cask Certificate of Compliance Format and Content

    Science.gov (United States)

    2013-02-05

    ... designers, major architect/engineering firms, and other organizations and entities involved in the nuclear... NRC-certified dry storage cask designs.'' III. The Petition In its petition (ADAMS Accession No... encompass the evaluation of the site-specific parameters versus the cask design bases information''...

  5. Regulation of dopamine release by CASK-β modulates locomotor initiation in Drosophila melanogaster

    Directory of Open Access Journals (Sweden)

    Justin eSlawson

    2014-11-01

    Full Text Available CASK is an evolutionarily conserved scaffolding protein that has roles in many cell types. In Drosophila, loss of the entire CASK gene or just the CASK-β transcript causes a complex set of adult locomotor defects. In this study, we show that the motor initiation component of this phenotype is due to loss of CASK-β in dopaminergic neurons and can be specifically rescued by expression of CASK-β within this subset of neurons. Functional imaging demonstrates that mutation of CASK-β disrupts coupling of neuronal activity to vesicle fusion. Consistent with this, locomotor initiation can be rescued by artificially driving activity in dopaminergic neurons. The molecular mechanism underlying this role of CASK-β in dopaminergic neurons involves interaction with Hsc70-4, a molecular chaperone previously shown to regulate calcium-dependent vesicle fusion. These data suggest that there is a novel CASK-β-dependent regulatory complex in dopaminergic neurons that serves to link activity and neurotransmitter release.

  6. Phenotypic and molecular insights into CASK-related disorders in males

    NARCIS (Netherlands)

    Moog, U.; Bierhals, T.; Brand, K.; Bautsch, J.; Biskup, S.; Brune, T.; Denecke, J.; Die-Smulders, C.E.M. de; Evers, C.; Hempel, M.; Henneke, M.; Yntema, H.G.; Menten, B.; Pietz, J.; Pfundt, R.P.; Schmidtke, J.; Steinemann, D.; Stumpel, C.T.; Maldergem, L. Van; Kutsche, K.

    2015-01-01

    BACKGROUND: Heterozygous loss-of-function mutations in the X-linked CASK gene cause progressive microcephaly with pontine and cerebellar hypoplasia (MICPCH) and severe intellectual disability (ID) in females. Different CASK mutations have also been reported in males. The associated phenotypes range

  7. 77 FR 24585 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-04-25

    ... 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... amends the NRC's spent fuel storage regulations by revising the Holtec International HI-STORM 100 System... International HI-STORM 100 System listing within the ``List of Approved Spent Fuel Storage Casks'' to...

  8. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Harrigan, R.W.; Sanders, T.L.

    1990-06-01

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs.

  9. Safety analysis report for packaging (onsite) multicanister overpack cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  10. Documentation for initial testing and inspections of Beneficial Uses Shipping System (BUSS) Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lundeen, J.E.

    1994-08-25

    The purpose of this report is to compile data generated during the initial tests and inspections of the Beneficial Uses Shipping System (BUSS) Cask. In addition, this report will verify that the testing criteria identified in section 8.1 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The BUSS Cask body and lid are each one-piece forgings fabricated from ASTM A473, Type 304 stainless steel. The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Chapter 8 of the BUSS Cask SARP requires several acceptance tests and inspections, each intended to evaluate the performance of different components of the BUSS Cask system, to be performed before its first use. The results of the tests and inspections required are included in this document.

  11. Implications of the Baltimore Rail Tunnel Fire for Full-Scale Testing of Shipping Casks

    Energy Technology Data Exchange (ETDEWEB)

    Halstead, R. J.; Dilger, F.

    2003-02-25

    The U.S. Nuclear Regulatory Commission (NRC) does not currently require full-scale physical testing of shipping casks as part of its certification process. Stakeholders have long urged NRC to require full-scale testing as part of certification. NRC is currently preparing a full-scale casktesting proposal as part of the Package Performance Study (PPS) that grew out of the NRC reexamination of the Modal Study. The State of Nevada and Clark County remain committed to the position that demonstration testing would not be an acceptable substitute for a combination of full-scale testing, scale-model tests, and computer simulation of each new cask design prior to certification. Based on previous analyses of cask testing issues, and on preliminary findings regarding the July 2001 Baltimore rail tunnel fire, the authors recommend that NRC prioritize extra-regulatory thermal testing of a large rail cask and the GA-4 truck cask under the PPS. The specific fire conditions and other aspects of the full-scale extra-regulatory tests recommended for the PPS are yet to be determined. NRC, in consultation with stakeholders, must consider past real-world accidents and computer simulations to establish temperature failure thresholds for cask containment and fuel cladding. The cost of extra-regulatory thermal testing is yet to be determined. The minimum cost for regulatory thermal testing of a legal-weight truck cask would likely be $3.3-3.8 million.

  12. Neuron-specific protein interactions of Drosophila CASK-b are revealed by mass spectrometry

    Directory of Open Access Journals (Sweden)

    Konark eMukherjee

    2014-06-01

    Full Text Available Modular scaffolding proteins are designed to have multiple interactors. CASK, a member of the membrane-associated guanylate kinase (MAGUK superfamily, has been shown to have roles in many tissues, including neurons and epithelia. It is likely that the set of proteins it interacts with is different in each of these diverse tissues. In this study we asked if within the Drosophila central nervous system, there were neuron-specific sets of CASK-interacting proteins. A YFP-tagged CASK transgene was expressed in genetically defined subsets of neurons in the Drosophila brain known to be important for CASK function, and proteins present in an anti-GFP immunoprecipitation were identified by mass spectrometry. Each subset of neurons had a distinct set of interacting proteins, suggesting that CASK participates in multiple protein networks and that these networks may be different in different neuronal circuits. One common set of proteins was associated with mitochondria, and we show here that endogenous CASK co-purifies with mitochondria. We also determined CASK posttranslational modifications for one cell type, supporting the idea that this technique can be used to assess cell- and circuit-specific protein modifications as well as protein interaction networks.

  13. Transfer cask system design activities: status and plan

    Energy Technology Data Exchange (ETDEWEB)

    Locke, D., E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Gutierrez, C. Gonzalez; Damiani, C.; Gracia, V. [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, J.-P.; Martins, J.-P.; Blight, J. [ITER Organisation, CS 90 046, 13067St. Paul Lez Durance Cedex (France)

    2011-10-15

    The ITER Cask and Plug Remote Handling System (CPRHS), a.k.a. Transfer Cask System, is a critical element of the ITER Remote Maintenance System (IRMS) devoted to transportation of components between the Tokamak building and Hot Cell. Due to the necessary confinement of contaminated components the CPRHS is defined as Safety Importance Class 1 (SIC-1) plus the mobile nature of the CPRHS brings with it a significant number of complex interfaces with other ITER sub-systems. With a total CPRHS fleet in excess of 20 units, including seven typologies, the management of design and procurement needs to be carefully planned and implemented to ensure compliance with ITER's requirements. Fusion for Energy (F4E) and its beneficiaries/contractors are currently working under ITER Task Agreements (ITAs) on the conceptual design of the CPRHS and, following the signing of the Procurement Arrangement (PA) in mid 2012, will take responsibility for the entire CPRHS fleet. F4E must, therefore, develop a robust strategy to meet the needs of both ITER machine assembly (for which a number of CPRHS units will be utilised) and the remote maintenance of ITER. Within this context this paper will present the status of the current CPRHS design activities, highlight some of the significant issues which will be faced during procurement and present the overall strategy which is being implemented by F4E in order to meet these challenging objectives.

  14. Contamination transfers during fuel transport cask loading. A concrete situation

    Energy Technology Data Exchange (ETDEWEB)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G. [DEN/DED Centre d' Etudes de Cadarache, 13 - Saint Paul lez Durance (France); Briquet, L. [EDF GENV, 93 - Saint Denis (France); Baubet, D. [SGS Qualitest Industrie, 30 - Pont Saint Esprit (France)

    2002-07-01

    In 1998, a number of contamination cases detected during fuel shipments have been pointed out by the french nuclear safety authority. Wagon and casks external surfaces were partly contaminated upon arrival in Valognes railway terminal. Since then, measures taken by nuclear power plants operators in France and abroad solved the problem. In Germany, a report analyzing the situation in depth has been published in which correctives actions have been listed. In France, EDF launched a large cleanliness program (projet proprete radiologique) in order to better understand contamination transfers mechanisms during power plants exploitation and to list remediation actions to avoid further problems. In this context, CEA Department for Wastes Studies at Cadarache (CEA/DEN/DED) was in charge of a study about contamination transfers during fuel elements loading operations. It was decided to lead experiments for a concrete case. The loading of a transport cask at Tricastin-PWR-1 was followed in november 2000 and different analysis comprising water analysis and smear tests analysis were carried out and are detailed in this paper. Results are discussed and qualitatively compared to those obtained in Philippsburg-BWR, Germany for a similar set of tests. (authors)

  15. CASKET: a computer code system for thermal and structural analyses of radioactive material transport and/or storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-05-01

    A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)

  16. Tandem SAM Domain Structure of Human Caskin1: A Presynaptic, Self-Assembling Scaffold for CASK

    Energy Technology Data Exchange (ETDEWEB)

    Stafford, Ryan L.; Hinde, Elizabeth; Knight, Mary Jane; Pennella, Mario A.; Ear, Jason; Digman, Michelle A.; Gratton, Enrico; Bowie, James U. (UCI); (UCLA)

    2012-02-07

    The synaptic scaffolding proteins CASK and Caskin1 are part of the fibrous mesh of proteins that organize the active zones of neural synapses. CASK binds to a region of Caskin1 called the CASK interaction domain (CID). Adjacent to the CID, Caskin1 contains two tandem sterile a motif (SAM) domains. Many SAM domains form polymers so they are good candidates for forming the fibrous structures seen in the active zone. We show here that the SAM domains of Caskin1 form a new type of SAM helical polymer. The Caskin1 polymer interface exhibits a remarkable segregation of charged residues, resulting in a high sensitivity to ionic strength in vitro. The Caskin1 polymers can be decorated with CASK proteins, illustrating how these proteins may work together to organize the cytomatrix in active zones.

  17. A method for determining the spent-fuel contribution to transport cask containment requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  18. Evaluation of computer programs used for structural analyses of impact response of spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B A; Gwinn, K W

    1984-05-01

    This report presents the results of a study of impact analyses of a generic spent-fuel cask. The study compares the use and results of three different finite element computer codes. Seven different cask-like model analyses are considered. The models encompass both linear and nonlinear geometric and material behavior. On the basis of the analyses results, this report recommends what parameters are useful in the comparison of different structural finite element computer programs. 5 references, 36 figures, 11 tables.

  19. NRC approves spent-fuel cask for general use: Who needs Yucca Mountain?

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, J.

    1993-07-01

    The Nuclear Regulatory Commission (NRC) on April 7, 1993, added Pacific Sierra Nuclear Associates`s (PSNA`s) VSC-24 spent-fuel container to its list of approved storage casks. Unlike previously approved designs, however, the cask was made available for use by utilities without site-specific approval. The VSC-24 (ventilated storage cask) is a 130-ton, 16-foot high vertical storage container composed of a ventilated concrete cask (VCC) housing a steel multi-assembly sealed basket (MSB). A third component, a transfer cask (MTC), shields, supports, and protects the MSB during fuel loading and VCC loading operations. The VCC is a cylindrical reinforced-concrete cask 29 inches thick, with a 1.75-inch-thick A 36 steel liner. The cask contains eight vents-four on the top and four on the bottom-to provide for MSB (and fuel rod) cooling. Its concrete shell provides protection against shearing and penetration by tornado projectiles, protects the MSB in the event of a drop or tipover, and is designed to withstand internal temperatures of 350 degrees Farenheit. The VCC is closed with a bolted-down cover of 0.75-inch-thick A 36 steel. The MSB, which provides the primary boundary for 24 spent fuel rods, is a cylindrical steel shell with a thick shield plug and steel cover plates welded at each end. The shell and covers are constructed from SA 516 Grade 70 pressure vessel steel. Fuel is housed in a basket fabricated from SA 516 Grade 70 sheet steel. Penetrations in the MSB`s structural and shield lids allow for vacuum drying and backfilling with helium after fuel loading. Although its manufacturer claims a design life of 50 years, the NRC has licensed the VSC-24 cask for 20 years.

  20. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  1. Operations manual for the Beneficial Uses Shipping System cask. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Bronowski, D.R.; Yoshimura, H.R.

    1993-04-01

    This document is the Operations Manual for the Beneficial Uses Shipping System (BUSS) cask. These operating instructions address requirements; for loading, shipping, and unloading, supplementing general operational information found in the BUSS Safety Analysis Report for Packaging (SARP), SAND 83-0698. Use of the BUSS cask is authorized by Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) for the shipment of special form cesium chloride or strontium flouride capsules.

  2. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il' kaev, R.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)

    2004-07-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks.

  3. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User`s manual to Version 3a. Volume 1, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L. [Lawrence Livermore National Lab., CA (United States)

    1998-03-01

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978.

  4. Cooling Performance Evaluation of the Hybrid Heat Pipe for Spent Nuclear Fuel Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    To evaluate the concept of the cooling device, 2-step CFD analysis was conducted for the cooling performance of hybrid heat pipe, which consists of single fuel assembly model and full scope dry cask model. As a passive cooling device of the metal cask for dry storage of spent nuclear fuel, hybrid heat pipe was applied to DPC developed in Korea. Hybrid heat pipe is the heat pipe containing neutron absorber can be used as a passive cooling in nuclear application with both decay heat removal and control the reactivity. In this study, 2-step CFD analysis was performed to find to evaluate the heat pipe-based passive cooling system for the application to the dry cask. Only spent fuel pool cannot satisfy the demands for high burnup fuel and large amount of spent fuel. Therefore, it is necessary to prepare supplement of the storage facilities. As one of the candidate of another type of storage, dry storage method have been preferred due to its good expansibility of storage capacity and easy long-term management. Dry storage uses the gas or air as coolant with passive cooling and neutron shielding materials was used instead of water in wet storage system. It is relatively safe and emits little radioactive waste for the storage. As short term actions for the limited storage capacity of spent fuel pool, it is considered to use dry interim/long term storage method to increase the capacity of spent nuclear fuel storage facilities. For 10-year cooled down spent fuel in the pool storage, fuel rod temperature inside metal cask is expected over 250 .deg. C in simulation. Although it satisfied the criteria that cladding temperature of the spent fuel should keep under 400 .deg. C during storage period, high temperature inside cask can accelerate the thermal degradation of the structural materials consisting metal cask and fuel assembly as well as limitation of the storage capacity of metal cask. In this paper, heat pipe-based cooling device for the dry storage cask was suggested for

  5. Conceptual Design Report Cask Loadout Sys and Cask Drop Redesign for the Immersion Pail Support Structure and Operator Interface Platform at 105 K West

    Energy Technology Data Exchange (ETDEWEB)

    LANGEVIN, A.S.

    1999-07-12

    This conceptual design report documents the redesign of the IPSS and the OIP in the 105 KW Basin south loadout pit due to a postulated cask drop accident, as part of Project A.5/A.6, Canister Transfer Facility Modifications. Project A.5/A.6 involves facility modifications needed to transfer fuel from the basin into the cask-MCO. The function of the IPSS is to suspend, guide, and position the immersion pail. The immersion pail protects the cask-MCO from contamination by basin water and acts as a lifting device for the cask-MCO. The OIP provides operator access to the south loadout pit. Previous analyses studied the effects of a cask-MCO drop on the south loadout pit concrete structure and on the IPSS. The most recent analysis considered the resulting loads at the pit slab/wall joint (Kanjilal, 1999). This area had not been modeled previously, and the analysis results indicate that the demand capacity exceeds the allowable at the slab/wall joint. The energy induced on the south loadout pit must be limited such that the safety class function of the basin is maintained. The solution presented in this CDR redesigns the IPSS and the OIP to include impact-absorbing features that will reduce the induced energy. The impact absorbing features of the new design include: Impact-absorbing material at the IPSS base and at the upper portion of the IPSS legs. A sleeve which provides a hydraulic means of absorbing energy. Designing the OIP to act as an impact absorber. The existing IPSS structure in 105 KW will be removed. This conceptual design considers only loads resulting from drops directly over the IPSS and south loadout pit area. Drops in other areas of the basin are not considered, and will be covered as part of a future revision to this CDR.

  6. Safety analysis report vitrified high level waste type B shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

  7. Structural evaluation and analysis under normal conditions for spent fuel concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Taechul; Baeg, Changyeal; Yoon, Sitae [Korea Radioactive waste Management Agency, Daejeon (Korea, Republic of); Jung, Insoo [Korea Nuclear Engineering and Service Co., Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this paper is the verification of stabilities of the structural elements that influence the safety of a concrete storage cask. The evaluation results were reviewed with respect to every design criterion, in terms of whether the results satisfy the criteria, provided by 10CFR 72 and NUREG-1536. The basic information on the design is partially explained in 2. Description of spent fuel storage system and the maintainability and assumptions included in the analysis were confirmed through detailed explanations of the acceptable standards, analysis model, and analysis method. ABAQUS 6.10, a widely used finite element analysis program, was used in the structural analysis. The storage cask shall maintain the sub-criticality, shielding, structural integrity, thermal capability and confinement in accordance with the requirements specified in US 10 CFR 72. The safety of storage cask is analyzed and it has been confirmed to meet the requirements of US 10 CFR 72. This paper summarizes the structural stability evaluation results of a concrete storage cask with respect to the design criteria. The evaluation results of this paper show that the maximum stress was below the allowable stress under every condition, and the concrete storage cask satisfied the design criteria.

  8. Documentation for fiscal year 1995 annual BUSS cask SARP testing and inspections

    Energy Technology Data Exchange (ETDEWEB)

    Saueressig, P.T.

    1994-11-08

    The purpose of this report is to compile the data generated during the Fiscal Year (FY) 1995 annual tests and inspections performed on the Beneficial Uses Shipping System (BUSS) cask. The BUSS Cask Model R-1 is a type B shipping container used for shipment of radioactive cesium-137 and strontium-90 capsules to Waste Encapsulation and Storage Facility (WESF). The primary purpose of the BUSS Cask is to provide shielding and confinement as well as impact, puncture, and thermal protection for the capsules under both normal and accident conditions. Section 8.2 ``Maintenance and Periodic Inspection Program`` of the BUSS Cask SARP requires that the following tests and inspections be performed on an annual basis: hydrostatic pressure test; helium leak test; dye penetrant test on the trunnions and life lugs; torque test on all permanent bolts; and impact limiter inspection and weight test. In addition to compiling the generated data, this report will verify that the testing criteria identified in section 8.2 of the BUSS Cask Safety Analysis Report for Packaging (SARP) was met.

  9. A Structural Analytic Evaluation of a Connote Pad In a Spent Fuel Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Hak; Seo, Ki Seog; Lee, Ju Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yeon Do; Cho, Chun Hyung; Lee, Dae Ki [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2006-06-15

    A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

  10. Study of minimum-weight highway transporters for spent nuclear fuel casks: Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Hoess, J.A.; Drago, V.J.

    1989-05-01

    There are federal and state limits on the maximum tractor-trailer- payload combination and individual axle loads permissible on US highways. These can generally be considered as two sets, i.e., legal-weight and overweight limits. The number of individual shipments required will decrease as the capacity of the spent nuclear fuel cask increases. Thus, there is an incentive for identifying readily available minimum-weight tractors and trailers capable of safely and reliably transporting as large a cask as possible without exceeding the legal gross combination weight (GCW) of 80,000 lb or selected overweight GCW limit of 110,000 lb. This study identifies options for commercially available heavy-duty on-highway tractors and trailers for transporting proposed future loaded spent nuclear fuel casks. Loaded cask weights of 56,000 and 80,000 lb were selected as reference design points for the legal-weight and overweight transporters, respectively. The technical data on tractor and trailer characteristics obtained indicate that it is possible to develop a tractor-trailer combination, tailored for spent nuclear fuel transportation service, utilizing existing technology and commercially available components, capable of safely and reliably transporting 56,000 and 80,000-lb spent nuclear fuel casks without exceeding GCWs of 80,000 and 10,000 lb, respectively. 4 figs., 14 tabs.

  11. A methodology for estimating the residual contamination contribution to the source term in a spent-fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L. (Sandia National Labs., Albuquerque, NM (United States)); Jordan, H. (EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant); Pasupathi, V. (Battelle, Columbus, OH (United States)); Mings, W.J. (USDOE Idaho Field Office, Idaho Falls, ID (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1991-09-01

    This report describes the ranges of the residual contamination that may build up in spent-fuel transport casks. These contamination ranges are calculated based on data taken from published reports and from previously unpublished data supplied by cask transporters. The data involve dose rate measurements, interior smear surveys, and analyses of water flushed out of cask cavities during decontamination operations. A methodology has been developed to estimate the effect of residual contamination on spent-fuel cask containment requirements. Factors in estimating the maximum permissible leak rates include the form of the residual contamination; possible release modes; internal gas-borne depletion; and the temperature, pressure, and vibration characteristics of the cask during transport under normal and accident conditions. 12 refs., 9 figs., 4 tabs.

  12. Evaluation of Impact Resistance of Concrete Overpack of Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The concrete overpack of the cask provides radiation shielding as well as physical protection for inner canister against external mechanical shock. When the overpack undergoes a severe missile impact which might be caused by tornado or aircraft crash, it should sustain minimal level of structural integrity so that the radiation shielding and the retrievability of canister are maintained. Empirical formulas have been developed for the evaluation of concrete damage but those formulas can be used only for local damage evaluation and not for global damage evaluation. In this research, a series of numerical simulations and tests have been performed to evaluate the damage of two types of concrete overpack segment models under high speed missile impact. It is shown that appropriate modeling of material failure is crucial in this kind of analyses and finding the correct failure parameters may not be straightforward. When comparing the simulation results with the test results, it is shown that neither setting, case 1 and 2 provides results with consistent agreement with test results. That is, case 1 setting is more close to reality in Type 1 model analysis, but for Type 2, case 2 setting provides more close results to the reality. In both the case, not enough deformation is predicted by simulation compared to the tests. Weak failure and eroding criteria give larger penetration depth with insufficient overall damage due to energy loss with element erosion.

  13. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    Science.gov (United States)

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.

  14. Development of Aircraft Impact Scenario on a Concrete Cask in Interim Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Momani, Belal Al; Yoo, Min; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    This paper provides a method for determining the failure criteria in global and local damage responses for the concrete cask under extreme mechanical impact condition. IAEA safety guide No. SSG-15 mentions the hypothetical initiating events of SNF storage. Among the external initiating events, the aircraft strike on a storage cask is considered one of the dominant contributions to the risk during storage phase. Although the probability of aircraft crash on ISF is extremely small, it is important to develop the accident scenario caused by an intentional malicious acts launched towards the storage facility in terms to improve inherent security. Thus, the probabilistic approach to develop aircraft impact scenarios on a storage cask is needed.

  15. Structural design of concrete storage pads for spent-fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Nickell, R.E.; James, R.J. (ANATECH Research Corp., San Diego, CA (United States))

    1993-04-01

    The loading experienced by spent fuel dry storage casks and storage pads due to potential drop or tip-over accidents is evaluated using state-of-the-art concrete structural analysis methodology. The purpose of this analysis is to provide simple design charts and formulas so that design adequacy of storage pads and dry storage casks can be demonstrated. The analysis covers a wide range of slab-design parameters, e.g., reinforcement ratio, slab thickness, concrete compressive strength, and sub-base soil compaction, as well as variations in drop orientation and drop height. The results are presented in the form of curves, giving the force on the cask as a function of storage pad hardness for various drop heights. In addition, force-displacement curves, deformed shapes, crack patterns, stresses and strains are given for various slab-design conditions and drop events. The utility of the results in design are illustrated through examples.

  16. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  17. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  18. TMI-2 (Three-Mile Island-Unit 2) rail cask and railcar maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.J.; Ayers, A.L., Jr.; Ball, L.J.; Anselmo, A.A.

    1988-02-01

    This paper describes the NuPac 125-B cask system (i.e., cask and railcar), and the maintenance and inspection requirements for that system. The paper discusses the operations being done to satisfy those requirements and how, in some cases, it has been efficient for the operations to be more rigorous than the requirements. Finally, this paper discusses the experiences gained from those operations and how specific hardware and procedural enhancements have resulted in a reliable and continuous shipping campaign. 2 refs., 2 figs.

  19. Ageing of a neutron shielding used in transport/storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique [TN International, 1 rue des herons, Montigny le Bretonneux, 78054 Saint Quentin en Yvelines (France); Laboratoire PIMM, Arts and Metiers ParisTech, 151 Bd de l' Hopital, 75013 Paris (France)

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  20. A structural analysis on the KN-12 spent nuclear fuel transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-08-15

    In this study, safety of the spent nuclear fuel cask KN-12 which is developed in 2000 is evaluated for hypothetical accidents conditions such as free drop, puncture, fire accident and water immersion. Finite element code ABAQUS/Explicit is used to compare with safety analysis report of the GNB in which analysis is performed with LS-DYNA3D for hypothetical accident conditions. Through this study, the safety of KN-12 is evaluated by comprehensive structural analysis. The capability and technological advancement of Korean community on the analysis and structural assessment of the cask will be improved. Also people's anxiety about radioactive dangers will be eliminated.

  1. Calcium/calmodulin-dependent serine protein kinase (CASK), a protein implicated in mental retardation and autism-spectrum disorders, interacts with T-Brain-1 (TBR1) to control extinction of associative memory in male mice

    Science.gov (United States)

    Huang, Tzyy-Nan; Hsueh, Yi-Ping

    2017-01-01

    Background Human genetic studies have indicated that mutations in calcium/calmodulin-dependent serine protein kinase (CASK) result in X-linked mental retardation and autism-spectrum disorders. We aimed to establish a mouse model to study how Cask regulates mental ability. Methods Because Cask encodes a multidomain scaffold protein, a possible strategy to dissect how CASK regulates mental ability and cognition is to disrupt specific protein–protein interactions of CASK in vivo and then investigate the impact of individual specific protein interactions. Previous in vitro analyses indicated that a rat CASK T724A mutation reduces the interaction between CASK and T-brain-1 (TBR1) in transfected COS cells. Because TBR1 is critical for glutamate receptor, ionotropic, N-methyl-d-aspartate receptor subunit 2B (Grin2b) expression and is a causative gene for autism and intellectual disability, we then generated CASK T740A (corresponding to rat CASK T724A) mutant mice using a gene-targeting approach. Immunoblotting, coimmunoprecipitation, histological methods and behavioural assays (including home cage, open field, auditory and contextual fear conditioning and conditioned taste aversion) were applied to investigate expression of CASK and its related proteins, the protein–protein interactions of CASK, and anatomic and behavioural features of CASK T740A mice. Results The CASK T740A mutation attenuated the interaction between CASK and TBR1 in the brain. However, CASK T740A mice were generally healthy, without obvious defects in brain morphology. The most dramatic defect among the mutant mice was in extinction of associative memory, though acquisition was normal. Limitations The functions of other CASK protein interactions cannot be addressed using CASK T740A mice. Conclusion Disruption of the CASK and TBR1 interaction impairs extinction, suggesting the involvement of CASK in cognitive flexibility. PMID:28234597

  2. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR... storage cask must be designed to provide adequate heat removal capacity without active cooling systems. (g... ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could...

  3. A Stylistic Analysis on Edgar Allan Poe's The Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    杨赛菲

    2016-01-01

    The Cask of Amontillado is one of Poe's best-known horror short stories. Based on Stylistics, this paper attempts to analyze this story from the aspects of themes, characterization, point of view, syntactic and lexical features, to reveal Poe's excellent skills and the artistic charm.

  4. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Buyun, E-mail: ayun@iim.ac.cn [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Gao, Lifu [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Cao, Huibin; Sun, Jian [Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Sun, Yuxiang; Song, Quanjun; Ma, Chengxue; Chang, Li; Shuang, Feng [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2014-12-15

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.

  5. ITER Transfer Cask System: Status of design, issues and future developments

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez, C. Gonzalez, E-mail: carmen.gonzalez@f4e.europa.e [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Damiani, C.; Irving, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, J.-P.; Tesini, A. [ITER Organization (IO), CS 90 046, 13067 St Paul Lez Durance Cedex (France); Ribeiro, I.; Vale, A. [Instituto Superior Tecnico (IST), ISR and IPFN, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2010-12-15

    The Remote Handling tasks scheduled during the ITER maintenance shutdown require transportation of in-vessel components from the Vacuum Vessel ports, at all levels of the Tokamak building, to the docking stations in the Hot Cell building. This transfer of radioactive, contaminated components represents a task of unprecedented complexity for a nuclear device like ITER. A Transfer Cask System (TCS) has been adopted as a reference solution. The TCS is a mobile, leak-tight unit, which can be divided into: (1) the cask itself, i.e., the container for the components to be transferred, able to avoid spread of contamination outside its envelope, equipped with in-cask handling devices; (2) the interface pallet that assists the docking operations of the cask and, underneath; (3) an Air Transfer System (ATS), i.e., a mobile platform floating on air-cushions with drive and steering wheels powered by electric motors and batteries on-board. The system will be remotely controlled, moving along previously defined paths. This paper focuses on the present status of the ATS design, the issues to be faced and the future developments foreseen.

  6. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Holtec International HI-STORM 100... and safety will be adequately protected. This direct final rule revises the HI-STORM 100 listing in...

  7. Structural dimensioning of dual purpose cask prototype; Dimensionamento estrutural de prototipo de casco de duplo proposito

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Luiz Leite da; Mourao, Rogerio Pimenta; Lopes, Claudio Cunha [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: silvall@cdtn.br; mouraor@cdtn.br; ccl@cdtn.br

    2005-07-01

    The structural dimensioning of a Type B(U) dual purpose cask prototype is part of the scope of work of the Brazilian institute CDTN in the IAEA regional project involving Latin American countries which operate research reactors (Argentina, Brazil, Chile, Mexico and Peru). In order to meet the dimensional and operational characteristics of the reactor facilities in these countries, a maximum weight of 10.000 kgf and a maximum dimension of 1 m in at least one direction were set for the cask. With these design restrictions, the cask's payload is either 21 MTR or 78 TRIGA fuel elements. The cask's most important components are main body, primary and secondary lids, basket and impact limiters. The main body has a sandwich-like wall with internal and external layers made of AISI 304 stainless steel with lead in-between. The lead provides biological shielding. The primary lid is similarly layered, but in the axial direction. It is provided with a double system of metallic rings and has ports for pressurization, sampling and containment verification. The secondary lid has the main function of protecting the primary lid against mechanical impacts. The basket structure is basically a tube array reinforced by bottom plate, feet and spacers. Square tubes are used for MTR elements and circular tubes for TRIGA elements. Finally, the impact limiters are structures made of an external stainless steel thin covering and a filling made of the wood composite OSB - Oriented Strand Board. The prototype is provided with bottom and top impact limiters, which are attached to each other by means of four threaded rods. The limiters are not rigidly attached to the cask body. A half scale cask model was designed to be submitted to a testing program. As its volume scales down to 1:8, the model weight is 1,250 kgf. This paper presents the methodology for the preliminary structural dimensioning of the critical parameters of the cask prototype. Both normal conditions of operation and

  8. Thermal modeling of a vertical dry storage cask for used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jie, E-mail: jieli@anl.gov; Liu, Yung Y., E-mail: yyliu@anl.gov

    2016-05-15

    Graphical abstract: - Highlights: • Thermal performance of a 3-D vertical dry cask under various conditions has been numerically studied by using ANSYS/FLUENT code. • The simulation was validated by comparing the results against data obtained from the temperature measurements of a commercial cask. • The results indicated that the basket with higher thermal conductivity dissipates decay heat out of the cask more efficiently than that with a lower thermal conductivity (aluminum composite vs. stainless steel). A heavier cooling gas is also helpful to enhance heat transfer via enhanced natural convection (N{sub 2} vs. He). • Coolant release from the fuel canister results in temperature change of the canister external surfaces. The simulation shows that such a change is large enough and detectable, which can provide a mechanism for leak detection by continuously monitoring this temperature change at the top center of the canister surface. • Partial blockage of the cask air inlets affects the temperature profiles marginally for both the fuel canister and those components inside. In contrast, fully blocked air inlets will lead to remarkable increases of the component temperatures. - Abstract: Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 × 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. The effects of canister fill gas (helium or nitrogen), internal pressure (1–6 atm), and basket material (stainless steel or aluminum alloy) were studied to

  9. Second Annual Maintenance, Inspection, and Test Report for PAS-1 Cask Certification for Shipping Payload B

    Energy Technology Data Exchange (ETDEWEB)

    KELLY, D.J.

    2000-10-09

    The Nuclear Packaging, Inc. (NuPac), PAS-1 cask is required to undergo annual maintenance and inspections to retain certification in accordance with U.S. Department of Energy (DOE) Certificate of Compliance USA/9184B(U) (Appendix A). The packaging configuration being tested and maintained is the NuPac PAS-1 cask for Payload B. The intent of the maintenance and inspections is to ensure the packaging remains in unimpaired physical condition. Two casks, serial numbers 2162-026 and 2162-027, were maintained, inspected, and tested at the 306E Development, Fabrication, and Test Laboratory, located at the Hanford Site's 300 Area. Waste Management Federal Services, Inc. (WMFS), a subsidiary of GTS Duratek, was in charge of the maintenance and testing. Cogema Engineering Corporation (Cogema) directed the operations in the test facility. The maintenance, testing, and inspections were conducted successfully with both PAS-1 casks. The work conducted on the overpacks included weighing, gasket replacement, and plastic pipe plug and foam inspections. The work conducted on the secondary containment vessel (SCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, and a helium leak test. The work conducted on the primary containment vessel (PCV) consisted of visual inspection of the vessel and threaded parts (i.e., fasteners), visual inspection of sealing surfaces, replacement of O-ring seals, dimensional inspection of the vessel bottom, a helium leak test, and dye penetrant inspection of the welds. The vermiculite material used in the cask rack assembly was replaced.

  10. Impact Analyses and Tests of Metal Cask Considering Aircraft Engine Crash - 12308

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sanghoon; Choi, Woo-Seok; Kim, Ki-Young; Jeon, Je-Eon; Seo, Ki-Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-07-01

    The structural integrity of a dual purpose metal cask currently under development by the Korea Radioactive Waste Management Cooperation (KRMC) is evaluated through analyses and tests under a high-speed missile impact considering the targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from the literature. The missile impact velocity was set at 150 m/s, and two impact orientations were considered. A simplified missile simulating a commercial aircraft engine is designed from an impact load history curve provided in the literature. In the analyses, the focus is on the evaluation of the containment boundary integrity of the metal cask. The analyses results are compared with the results of tests using a 1/3 scale model. The results show very good agreements, and the procedure and methodology adopted in the structural analyses are validated. While the integrity of the cask is maintained in one evaluation where the missile impacts the top side of the free standing cask, the containment boundary is breached in another case in which the missile impacts the center of the cask lid in a perpendicular orientation. A safety assessment using a numerical simulation of an aircraft engine crash into spent nuclear fuel storage systems is performed. A commercially available explicit finite element code is utilized for the dynamic simulation, and the strain rate effect is included in the modeling of the materials used in the target system and missile. The simulation results show very good agreement with the test results. It is noted that this is the first test considering an aircraft crash in Korea. (authors)

  11. Metal cask RT-5000 for the dry storage and transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vorobyov, A.I.; Kazeev, V.G.; Krayev, V.S.; Shcherbina, A.N.; Churikov, Y.I. [All-Russian Research Inst. of Technical Physics, Snezhinsk (Russian Federation)

    2003-05-01

    Presentation of new-type cask, developed at RFNC-VNIITF, is in the article. The prototype model of the shipping cask was subjected to tests imitating normal shipment conditions (free fall, pressing, and impact) and to tests imitating emergency situation during shipment (a drop from the 9-m height onto a pin is replaced by acceleration of the shipping cask at a guide rail of the rocket-catapult installation (RCI), a 1-m drop onto a pin, heat tests a 30-minutes fire at the temperature of for 8500 C, submergence to the depth of 15 and 200 meters). After each test the hermeticity preservation is examined. Parallel with the real testing, a mathematical simulation of physical processes induced by the corresponding tests was conducted at the RFNC-VNIITF. The required parameters obtained from the tests are used to calibrate the calculation methods. As a result it has been possible to obtain a good agreement between the results of calculations and experiments; this will allow the mathematic simulation to be used wider. The advantage of the RT-5000 metal cask in comparison with metal-concrete analogs are as follows: SFA are placed into the RT-5000 entirely without cutting into two bunches of fuel elements; the expensive hot doom is not required for automatic cutting the SFA and for loading the bunches of fuel elements into intermediate cases; the possibility remains to transport the RT-5000 without reloading SFA after 50-year storage, although this is a problem for the metal-concrete casks.

  12. Thermo-mechanical finite element analyses of bolted cask lid structures

    Energy Technology Data Exchange (ETDEWEB)

    Wieser, G.; Qiao Linan; Eberle, A.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    The analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak-tightness in package design assessment according to the Transport Regulations or in aircraft crash scenarios. In this context BAM is developing methods based on Finite Elements to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. I n case of fire it might be not enough to perform only a thermal heat transfer analysis. The complex cask design in connection with a severe hypothetical time-temperature-curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and its counterparts that can be analyzed by a so-called thermo-mechanical calculation. Although it is currently not possible to correlate leakage rates with results from deformation analyses directly an appropriate Finite Element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field respectively. Except of the lid bolts the geometry of the cask and the mechanical loading is axial-symmetric which simplifies the analysis considerably and a two-dimensional Finite Element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modelling on the relative displacements at the seating of the seals. Besides this, the influence of bolt modelling, thermal properties and detail in geometry of the two-dimensional Finite Element models on

  13. PRELIMINARY REPORT: EFFECTS OF IRRADIATION AND THERMAL EXPOSURE ON ELASTOMERIC SEALS FOR CASK TRANSPORTATION AND STORAGE

    Energy Technology Data Exchange (ETDEWEB)

    Verst, C.; Skidmore, E.; Daugherty, W.

    2014-05-30

    A testing and analysis approach to predict the sealing behavior of elastomeric seal materials in dry storage casks and evaluate their ability to maintain a seal under thermal and radiation exposure conditions of extended storage and beyond was developed, and initial tests have been conducted. The initial tests evaluate the aging response of EPDM elastomer O-ring seals. The thermal and radiation exposure conditions of the CASTOR® V/21 casks were selected for testing as this cask design is of interest due to its widespread use, and close proximity of the seals to the fuel compared to other cask designs leading to a relatively high temperature and dose under storage conditions. A novel test fixture was developed to enable compression stress relaxation measurements for the seal material at the thermal and radiation exposure conditions. A loss of compression stress of 90% is suggested as the threshold at which sealing ability of an elastomeric seal would be lost. Previous studies have shown this value to be conservative to actual leakage failure for most aging conditions. These initial results indicate that the seal would be expected to retain sealing ability throughout extended storage at the cask design conditions, though longer exposure times are needed to validate this assumption. The high constant dose rate used in the testing is not prototypic of the decreasingly low dose rate that would occur under extended storage. The primary degradation mechanism of oxidation of polymeric compounds is highly dependent on temperature and time of exposure, and with radiation expected to exacerbate the oxidation.

  14. Behavior of spent fuel and cask components after extended periods of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Kenneally, R. [U.S. Nuclear Regulatory Commission, Rockville, MD (United States); Kessler, J. [Electric Power Research Inst., Palo Alto, CA (United States)

    2001-07-01

    The U.S. Nuclear Regulatory Commission (NRC) promulgated 10 CFR Part 72, Title 10, for the independent storage of spent nuclear fuel and high-level radioactive waste outside reactor spent fuel pools. Part 72 currently limits the license term for an independent spent fuel storage installation to 20 years from the date of issuance. Licenses may be renewed by the Commission at or before the expiration of the license term. Applications for renewal of a license should be filed at least two years prior to the expiration of the existing license. In preparation for possible license renewal, the NRC Office of Nuclear Material and Safeguards, Spent Fuel Project Office, is developing the technical basis for renewals of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level radioactive waste at independent spent fuel storage installation sites. An analysis of past performance of selected components of these systems is required as part of that technical basis. In the years 1980 through the early 1990, the Department of Energy (DOE) procured four prototype dry storage casks for testing at the Idaho National Engineering and Environmental Laboratory (INEEL): Castor-V/21, MC-10, TN-24P, and VSC-17. The primary purpose of the testing was to benchmark thermal and radiological codes and to determine the thermal and radiological characteristics of the casks. A series of examinations in 1999 and early 2000 to investigate the integrity of the Castor V/21 cask were undertaken. There is no evidence of significant degradation of the Castor V/21 cask systems important to safety from the time of initial loading of the cask in 1985 up to the time of testing in 1999. (author)

  15. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  16. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  17. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Ho Chul; Hong, Song Jin; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-03-15

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS is completely corresponded. And The integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

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

  19. Nonlinear Ultrasonic Diagnosis and Prognosis of ASR Damage in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Qu, Jianmin [Northwestern Univ., Evanston, IL (United States); Bazant, Zdenek [Northwestern Univ., Evanston, IL (United States); Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Guimaraes, Maria [Electrical Power Research Institute, Palo Alto, CA (United States)

    2015-11-30

    Alkali-silica reaction (ASR) is a deleterious chemical process that may occur in cement-based materials such as mortars and concretes, where the hydroxyl ions in the highly alkaline pore solution attack the siloxane groups in the siliceous minerals in the aggregates. The reaction produces a cross-linked alkali-silica gel. The ASR gel swells in the presence of water. Expansion of the gel results in cracking when the swelling-induced stress exceeds the fracture toughness of the concrete. As the ASR continues, cracks may grow and eventually coalesce, which results in reduced service life and a decrease safety of concrete structures. Since concrete is widely used as a critical structural component in dry cask storage of used nuclear fuels, ASR damage poses a significant threat to the sustainability of long term dry cask storage systems. Therefore, techniques for effectively detecting, managing and mitigating ASR damage are needed. Currently, there are no nondestructive methods to accurately detect ASR damage in existing concrete structures. The only current way of accurately assessing ASR damage is to drill a core from an existing structure, and conduct microscopy on this drilled cylindrical core. Clearly, such a practice is not applicable to dry cask storage systems. To meet these needs, this research is aimed at developing (1) a suite of nonlinear ultrasonic quantitative nondestructive evaluation (QNDE) techniques to characterize ASR damage, and (2) a physics-based model for ASR damage evolution using the QNDE data. Outcomes of this research will provide a nondestructive diagnostic tool to evaluate the extent of the ASR damage, and a prognostic tool to estimate the future reliability and safety of the concrete structures in dry cask storage systems

  20. Alternative Splicing of a Novel Inducible Exon Diversifies the CASK Guanylate Kinase Domain

    Directory of Open Access Journals (Sweden)

    Jill A. Dembowski

    2012-01-01

    Full Text Available Alternative pre-mRNA splicing has a major impact on cellular functions and development with the potential to fine-tune cellular localization, posttranslational modification, interaction properties, and expression levels of cognate proteins. The plasticity of regulation sets the stage for cells to adjust the relative levels of spliced mRNA isoforms in response to stress or stimulation. As part of an exon profiling analysis of mouse cortical neurons stimulated with high KCl to induce membrane depolarization, we detected a previously unrecognized exon (E24a of the CASK gene, which encodes for a conserved peptide insertion in the guanylate kinase interaction domain. Comparative sequence analysis shows that E24a appeared selectively in mammalian CASK genes as part of a >3,000 base pair intron insertion. We demonstrate that a combination of a naturally defective 5 splice site and negative regulation by several splicing factors, including SC35 (SRSF2 and ASF/SF2 (SRSF1, drives E24a skipping in most cell types. However, this negative regulation is countered with an observed increase in E24a inclusion after neuronal stimulation and NMDA receptor signaling. Taken together, E24a is typically a skipped exon, which awakens during neuronal stimulation with the potential to diversify the protein interaction properties of the CASK polypeptide.

  1. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

  2. Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio, E-mail: romanato@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade

    2011-07-01

    Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

  3. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  4. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, A.; Koploy, M.

    1991-12-01

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

  5. Dynamic Response Analysis of Storage Cask Lid Structure Subjected to Lateral Impact Load of Aircraft Engine Crash

    Energy Technology Data Exchange (ETDEWEB)

    Almomania, Belal; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Lee, Sanghoon [Keimyung Univ., Daegu (Korea, Republic of)

    2015-10-15

    Several numerical methods and tests have been carried out to measure the capability of storage cask to withstand extreme impact loads. Testing methods are often constrained by cost, and difficulty in preparation for several impact conditions with different applied loads, and areas of impact. Instead, analytic method is an acceptable process that can easily apply different impact conditions for the evaluation of cask integrity. The aircraft engine impact is considered as one of the most critical impact accidents on the storage cask that significantly affects onto the lid closure system and may cause a considerable release of radioactive materials. This paper presents a method for evaluating the dynamic responses of one upper metal cask lid closure without impact limiters subjected to lateral impact of an aircraft engine with respect to variation of the impact velocity. An assessment method to predict damage response due to the lateral engine impact onto metal storage cask has been studied by using computer code LS-DYNA. The dynamic behavior of the lid movements was successfully calculated by utilizing a simplified finite element cask model, which showed a good agreement with the previous research. The simulation analyses results showed that no significant plastic deformation for bolts, lid, and the cask body. In this study, the lid opening and sliding displacements are considered as the major factors in initiating the leakage path. This analysis may be useful for evaluating the instantaneous leakage rates in a connection with the sliding and opening displacements between the lid and the flange to ensure that the radiological consequences caused by an aircraft engine crash accident during the storage phase are within the permissible level.

  6. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    Energy Technology Data Exchange (ETDEWEB)

    Manson, S.J.; Gianoulakis, S.E. [Sandia National Labs., Albuquerque, NM (United States). Transportation Systems Development Dept.

    1994-04-01

    An examination of the effect of a realistic (though conservative) hot day environment on the thermal transient behavior of spent fuel shipping casks is made. These results are compared to those that develop under the prescribed normal thermal condition of 10 CFR 71. Of specific concern are the characteristics of propagating thermal waves, which are set up by diurnal variations of temperature and insolation in the outdoor environment. In order to arrive at a realistic approximation of these variations on a conservative hot day, actual temperature and insolation measurements have been obtained from the National Climatic Data Center (NCDC) for representatively hot and high heat flux days. Thus, the use of authentic meteorological data ensures the realistic approach sought. Further supporting the desired realism of the modeling effort is the use of realistic cask configurations in which multiple laminations of structural, shielding, and other materials are expected to attenuate the propagating thermal waves. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by enforcement of the regulatory environmental conditions of 10 CFR 71. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the prescribed regulatory conditions. However, the temperature differences are small enough that the normal conservative assumptions that are made in the course of typical cask evaluations should correct for any potential violations. The analysis demonstrates that diurnal temperature variations that penetrate the cask wall all have maxima substantially less than the corresponding regulatory solutions. Therefore it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the conditions of 10 CFR 71.

  7. Behaviour of neutron moderator materials at high temperatures in CASTOR {sup registered} -casks: qualification and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Krietsch, T.; Wolff, D. [Federal Inst. for Materials Research and Testing (BAM), Berlin (Germany); Knopp, U. [Gesellschaft fuer Nuklear-Behaelter mbH (GNB), Essen (Germany); Brocke, H.D. [TUeV Rheinland Group, Berlin (Germany)

    2004-07-01

    The Federal Institute for Materials Research and Testing (BAM) is the responsible German authority for the assessment of mechanical and thermal designs of transport and storage casks for radioactive materials. BAM checks up the proofs of the applicants in their safety reports and assesses the conformity to the Regulations for the Safe Transport of Radioactive Material. One applicant is the Gesellschaft fuer Nuklear-Behaelter mbH (GNB) with a new generation of transport and storage casks of CASTOR {sup registered} -design. GNB typically uses ultra high molecular weight Polyethylene (UHMW-PE) for the moderation of free neutrons. Rods made of UHMW-PE are positioned in axial bore holes in the wall of the cask and plates of UHMW-PE are in free spaces between primary and secondary lid and between the bottom of the cask and an outer plate (Figure 1). Because of the heat generated by the radioactive inventory and because of a strained spring at the bottom of every bore hole, UHMW-PE is subjected to permanent thermal and mechanical loads as well as loads from gamma and neutron radiation. UHMW-PE has been used under routine- and normal conditions of transport for maximum temperatures up to 130 C. For new generations of CASTOR registered -design maximum temperatures will be increased up to 160 C. That means a permanent use of UHMW-PE at temperatures within and above the melting region of the crystallites. In this paper, some results of special investigations for the proofs of usability of UHMW-PE at temperatures up to 160 C under real conditions of transport and storage in CASTOR registered -casks are given. For that, investigations on temperature dependent expansion behaviour under laboratory conditions as well as in large scale experiments, especially in the case of multiple heating and cooling, were done. Besides, geometrical creep strength for long-term loading by temperatures and pressures with regard to the chemical and physical stability properties of UHMW-PE above the

  8. Discussion of Available Methods to Support Reviews of Spent Fuel Storage Installation Cask Drop Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Witte, M.

    2000-03-28

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  9. CASK interacts with PMCA4b and JAM-A on the mouse sperm flagellum to regulate Ca2+ homeostasis and motility.

    Science.gov (United States)

    Aravindan, Rolands G; Fomin, Victor P; Naik, Ulhas P; Modelski, Mark J; Naik, Meghna U; Galileo, Deni S; Duncan, Randall L; Martin-Deleon, Patricia A

    2012-08-01

    Deletion of the highly conserved gene for the major Ca(2+) efflux pump, Plasma membrane calcium/calmodulin-dependent ATPase 4b (Pmca4b), in the mouse leads to loss of progressive and hyperactivated sperm motility and infertility. Here we first demonstrate that compared to wild-type (WT), Junctional adhesion molecule-A (Jam-A) null sperm, previously shown to have motility defects and an abnormal mitochondrial phenotype reminiscent of that seen in Pmca4b nulls, exhibit reduced (P JAM-A on the proximal principal piece, acts as a common interacting partner of both. Importantly, CASK binds alternatively and non-synergistically with each of these molecules via its single PDZ (PDS-95/Dlg/ZO-1) domain to either inhibit or promote efflux. In the absence of CASK-JAM-A interaction in Jam-A null sperm, CASK-PMCA4b interaction is increased, resulting in inhibition of PMCA4b's enzymatic activity, consequent Ca(2+) accumulation, and a ∼6-fold over-expression of constitutively ATP-utilizing CASK, compared to WT. Thus, CASK negatively regulates PMCA4b by directly binding to it and JAM-A positively regulates it indirectly through CASK. The decreased motility is likely due to the collateral net deficit in ATP observed in nulls. Our data indicate that Ca(2+) homeostasis in sperm is maintained by the relative ratios of CASK-PMCA4b and CASK-JAM-A interactions.

  10. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  11. Safety aspects of long-term dry interim storage of type-B spent fuel and HLW transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    Based on the German decision to minimise transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities have been licensed till the end of 2003. Because of the large amount of type-B transport casks which are going to be used for long-term interim storage the question of time limited type-B license maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of transport cask qualification for interim storage is the long-term behaviour of the metallic seal lid system. Concerning this results from current experimental long-term tests with metallic ''Helicoflex''-seals in which pool water is enclosed are presented. The test series has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an authorities' and technical experts' exchange of experience, know-how and state of the art referring to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called ''co-ordinating institution for cask dispatching information'' (''KOBAF'') which contains an online data base and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the next decades.

  12. Identification and glycerol-induced correction of misfolding mutations in the X-linked mental retardation gene CASK.

    Directory of Open Access Journals (Sweden)

    Leslie E W LaConte

    Full Text Available The overwhelming amount of available genomic sequence variation information demands a streamlined approach to examine known pathogenic mutations of any given protein. Here we seek to outline a strategy to easily classify pathogenic missense mutations that cause protein misfolding and are thus good candidates for chaperone-based therapeutic strategies, using previously identified mutations in the gene CASK. We applied a battery of bioinformatics algorithms designed to predict potential impact on protein structure to five pathogenic missense mutations in the protein CASK that have been shown to underlie pathologies ranging from X-linked mental retardation to autism spectrum disorder. A successful classification of the mutations as damaging was not consistently achieved despite the known pathogenicity. In addition to the bioinformatics analyses, we performed molecular modeling and phylogenetic comparisons. Finally, we developed a simple high-throughput imaging assay to measure the misfolding propensity of the CASK mutants in situ. Our data suggests that a phylogenetic analysis may be a robust method for predicting structurally damaging mutations in CASK. Mutations in two evolutionarily invariant residues (Y728C and W919R exhibited a strong propensity to misfold and form visible aggregates in the cytosolic milieu. The remaining mutations (R28L, Y268H, and P396S showed no evidence of aggregation and maintained their interactions with known CASK binding partners liprin-α3 Mint-1, and Veli, indicating an intact structure. Intriguingly, the protein aggregation caused by the Y728C and W919R mutations was reversed by treating the cells with a chemical chaperone (glycerol, providing a possible therapeutic strategy for treating structural mutations in CASK in the future.

  13. A novel interaction between FRMD7 and CASK: evidence for a causal role in idiopathic infantile nystagmus.

    Science.gov (United States)

    Watkins, Rachel J; Patil, Rajashree; Goult, Benjamin T; Thomas, Mervyn G; Gottlob, Irene; Shackleton, Sue

    2013-05-15

    Idiopathic infantile nystagmus (IIN) is a genetically heterogeneous disorder of eye movement that can be caused by mutations in the FRMD7 gene that encodes a FERM domain protein. FRMD7 is expressed in the brain and knock-down studies suggest it plays a role in neurite extension through modulation of the actin cytoskeleton, yet little is known about its precise molecular function and the effects of IIN mutations. Here, we studied four IIN-associated missense mutants and found them to have diverse effects on FRMD7 expression and cytoplasmic localization. The C271Y mutant accumulates in the nucleus, possibly due to disruption of a nuclear export sequence located downstream of the FERM-adjacent domain. While overexpression of wild-type FRMD7 promotes neurite outgrowth, mutants reduce this effect to differing degrees and the nuclear localizing C271Y mutant acts in a dominant-negative manner to inhibit neurite formation. To gain insight into FRMD7 molecular function, we used an IP-MS approach and identified the multi-domain plasma membrane scaffolding protein, CASK, as a FRMD7 interactor. Importantly, CASK promotes FRMD7 co-localization at the plasma membrane, where it enhances CASK-induced neurite length, whereas IIN-associated FRMD7 mutations impair all of these features. Mutations in CASK cause X-linked mental retardation. Patients with C-terminal CASK mutations also present with nystagmus and, strikingly, we show that these mutations specifically disrupt interaction with FRMD7. Together, our data strongly support a model whereby CASK recruits FRMD7 to the plasma membrane to promote neurite outgrowth during development of the oculomotor neural network and that defects in this interaction result in nystagmus.

  14. Interaction of cosmic ray muons with spent nuclear fuel dry casks and determination of lower detection limit

    Science.gov (United States)

    Chatzidakis, S.; Choi, C. K.; Tsoukalas, L. H.

    2016-08-01

    The potential non-proliferation monitoring of spent nuclear fuel sealed in dry casks interacting continuously with the naturally generated cosmic ray muons is investigated. Treatments on the muon RMS scattering angle by Moliere, Rossi-Greisen, Highland and, Lynch-Dahl were analyzed and compared with simplified Monte Carlo simulations. The Lynch-Dahl expression has the lowest error and appears to be appropriate when performing conceptual calculations for high-Z, thick targets such as dry casks. The GEANT4 Monte Carlo code was used to simulate dry casks with various fuel loadings and scattering variance estimates for each case were obtained. The scattering variance estimation was shown to be unbiased and using Chebyshev's inequality, it was found that 106 muons will provide estimates of the scattering variances that are within 1% of the true value at a 99% confidence level. These estimates were used as reference values to calculate scattering distributions and evaluate the asymptotic behavior for small variations on fuel loading. It is shown that the scattering distributions between a fully loaded dry cask and one with a fuel assembly missing initially overlap significantly but their distance eventually increases with increasing number of muons. One missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the distributions which is the case of 100,000 muons. This indicates that the removal of a standard fuel assembly can be identified using muons providing that enough muons are collected. A Bayesian algorithm was developed to classify dry casks and provide a decision rule that minimizes the risk of making an incorrect decision. The algorithm performance was evaluated and the lower detection limit was determined.

  15. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  16. Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    2001-03-08

    The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent fuel storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent nuclear fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS that define the acceptable fuel types. To support this initiative, this study has been performed to identify potential fuel specification parameters needed for criticality safety and radiation shielding analysis and rank their importance relative to a potential compromise of the margin of safety.

  17. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-12-01

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson`s ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user`s guide for computer program and input data for the IMPACLIB. (author)

  18. IMPACT ANALYSES AND TESTS OF CONCRETE OVERPACKS OF SPENT NUCLEAR FUEL STORAGE CASKS

    Directory of Open Access Journals (Sweden)

    SANGHOON LEE

    2014-02-01

    Full Text Available A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

  19. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  20. 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)

  1. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  2. Inspection and Gamma-Ray Dose Rate Measurements of the Annulus of the VSC-17 Concrete Spent Nuclear Fuel Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    P. L. Winston

    2007-09-01

    The air cooling annulus of the Ventilated Storage Cask (VSC)-17 spent fuel storage cask was inspected using a Toshiba 7 mm (1/4”) CCD video camera. The dose rates observed in the annular space were measured to provide a reference for the activity to which the camera(s) being tested were being exposed. No gross degradation, pitting, or general corrosion was observed.

  3. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, N.; Ishida, N.; Ootsuka, M.; Kamoshida, M.; Hiranuma, T.; Doumori, S.; Hoshikawa, T.; Shimizu, M.; Kashiwakura, J.; Hayashi, M. [Hitachi, Ltd., Hitachi (Japan)

    2004-07-01

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

  4. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kyoung-Sik, E-mail: nksbang@kaeri.re.kr; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-08-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  5. Estimates of durability of TMI-2 core debris canisters and cask liners

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Lund, A.L.; Pednekar, S.P.

    1994-04-01

    Core debris from the Three Mile Island-2 (TMI-2) reactor is currently stored in stainless steel canisters. The need to maintain the integrity of the TMI-2 core debris containers through the period of extended storage and possibly into disposal prompted this assessment. In the assessment, corrosion-induced degradation was estimated for two materials: type 304L stainless steel (SS) canisters that contain the core debris, and type 1020 carbon steel (CS) liners in the concrete casks planned for containing the canisters from 2000 AD until the TMI-2 core debris is placed in a repository. Three environments were considered: air-saturated water (with 2 ppM Cl{sup {minus}}) at 20{degree}C, and air at 20{degree}C with two relative humidities (RHs), 10 and 40%. Corrosion mechanisms assessed included general corrosion (failure criterion: 50% loss of wall thickness) and localized attack (failure criterion: through-wall pinhole penetration). Estimation of carbon steel corrosion after 50 y also was requested.

  6. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  7. Acoustic emission detection with fiber optical sensors for dry cask storage health monitoring

    Science.gov (United States)

    Lin, Bin; Bao, Jingjing; Yu, Lingyu; Giurgiutiu, Victor

    2016-04-01

    The increasing number, size, and complexity of nuclear facilities deployed worldwide are increasing the need to maintain readiness and develop innovative sensing materials to monitor important to safety structures (ITS). In the past two decades, an extensive sensor technology development has been used for structural health monitoring (SHM). Technologies for the diagnosis and prognosis of a nuclear system, such as dry cask storage system (DCSS), can improve verification of the health of the structure that can eventually reduce the likelihood of inadvertently failure of a component. Fiber optical sensors have emerged as one of the major SHM technologies developed particularly for temperature and strain measurements. This paper presents the development of optical equipment that is suitable for ultrasonic guided wave detection for active SHM in the MHz range. An experimental study of using fiber Bragg grating (FBG) as acoustic emission (AE) sensors was performed on steel blocks. FBG have the advantage of being durable, lightweight, and easily embeddable into composite structures as well as being immune to electromagnetic interference and optically multiplexed. The temperature effect on the FBG sensors was also studied. A multi-channel FBG system was developed and compared with piezoelectric based AE system. The paper ends with conclusions and suggestions for further work.

  8. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user`s guide for computer program and input data for THERMLIB. (author)

  9. Spent fuel criticality and compositions evaluation for long-term disposal in a generic cask

    Energy Technology Data Exchange (ETDEWEB)

    Velasquez, C.E.; Sousa, R.V.; Fortini, A.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Costa, A.L.; Silva, C.A.M. da; Veloso, M.A.F.; Oliveira, A.H. de; Carvalho, F.R. de

    2014-08-15

    The Nuclear Energy Agency (NEA) Expert Group on Burn-up Credit Criticality Safety published a Benchmark with results obtained from simulations with some nuclear codes for a PWR-UO{sub 2} nuclear fuel disposed of in a cask. The same situations were simulated at the Departamento de Engenharia Nuclear/Universidade Federal de Minas Gerais (DEN/UFMG) with the SCALE 6.0 (KENOVI/ORIGENS), MCNPX 2.6.0/CINDER and Monteburns (MCNP5/ORIGEN2.1). Combinations of codes and nuclear data are slightly different from those used by the organizations who participate of the Benchmark. For k{sub eff} time evolution, the results are very similar to the values obtained by the benchmark participants. For decay time evolution, the results obtained for several nuclides presented the expected behavior. Nevertheless, differences in the composition increase during the time specially using the Monteburns code. These differences may be attributed to the libraries and methodology for choosing libraries to decay calculation and the number of days to a year considered to calculations.

  10. Large deformation inelastic analysis of impact for shipping casks. [DYNA3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Charman, C.M.; Grenier, R.M. (General Atomic Co., San Diego, CA (USA)); Nickell, R.E. (Applied Science and Technology, Poway, CA (USA))

    1982-09-01

    This paper describes the use of two- and three-dimensional nonlinear finite element computer programs to design a radioative material transportation cask to withstand a drop of 30 feet onto an unyielding surface. Because of recent advancement in the area of non-linear finite element code development, the use of such codes for an iterative design process is becoming practicable. The paper begins with a section dealing with a two-dimensional side drop analysis and is followed by a discussion of the general capabilities of DYNA3D and a brief discussion of the implementation of the code on a computational mainframe unlike any for which the developer had intended. Then, a section on three-dimensional models of center-of-gravity over a corner impact follows, which introduces design features such as bolted closures, internal impact limiter, seals and shear rings. Figs. showing the deformed model grids are included. Stress and strain results are given in the subsequent section. Finally, we interpret these results in terms of possible rules being developed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code committees.

  11. Calculative activation analysis of the transport rack for CASTOR {sup registered} casks; Berechnung der Aktivierung eines Transportgestells fuer CASTOR {sup registered} -Behaelter

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany); Biedermann, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Schmidt-Wohlfarth, Y.; Louia, A. [EnBW Kernkraft GmbH, Philippsburg (Germany)

    2011-07-01

    The transport rack for the internal transport of loaded CASTOR {sup registered} casks before the storage in the intermediate storage facility at the site of the NPP Philippsburg is exposed to neutron irradiation from the cask inventory. Using the Monte Carlo code MCNP the activation rates of the transport rack materials are calculated for typical storage times of the casks in the rack. The long-term activation was also calculated for the continuous use of the transport rack over 10 years. Further topics were the dose rate in the near surrounding of the transport rack after long-term activation and finally the disposability of rack components according to the legal regulations. The maximum contact dose rate was calculated to be below 1 micro Sv/h after 10 years of application. The transport rack can be disposed with large safety margins to the radiation protection limits.

  12. CAPSIZE: A personal computer program and cross-section library for determining the shielding requirements, size, and capacity of shipping casks subject to various proposed objectives

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1987-05-01

    A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs.

  13. Shielding Research of Cobalt Adjuster Rod Transport Cask%钴调节棒转运容器屏蔽研究

    Institute of Scientific and Technical Information of China (English)

    王炳衡; 薛娜; 毛亚蔚

    2013-01-01

    Dose rates at the surface of adjuster rod transport cask were calculated with MCNP code and MCAM code.The calculation results were compared with national standard to estimate shielding design of the cask.To meet the requirement of the standard,original design of the cask was updated through the calculation of these codes.To reduce the radiation level,temporary shielding for the gap (between shielding door and main cask) was applied and 5 cm shielding layer of depleted uranium was applied to the side shield instead of 5 cm lead layer.According to the results,new shielding design of transport cask can meet the national standard.The measuring results show that the design of the cask is reasonable and reliable.%采用MCAM程序与MCNP程序模拟计算钴调节棒转运容器的表面剂量率,并以此来判断容器的屏蔽设计是否满足标准要求.通过程序系统估算,在容器初始设计模型的基础上将5 cm铅层替换为5cm贫铀防护层,并提出了在容器下部屏蔽门缝隙处增加临时屏蔽装置以降低该处的辐射水平.经过优化设计后,钴调节棒转运容器能够满足国家相应的屏蔽标准要求.现场操作时的实测结果也进一步验证了容器屏蔽设计的合理性和可靠性.

  14. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    Energy Technology Data Exchange (ETDEWEB)

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures.

  15. Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsoukalas, Lefteri H. [Purdue Univ., West Lafayette, IN (United States)

    2016-03-01

    This is the final report of the NEUP project “Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage”, DE-NE0000695. The project started on December 1, 2013 and this report covers the period December 1, 2013 through November 30, 2015. The project was successfully completed and this report provides an overview of the main achievements, results and findings throughout the duration of the project. Additional details can be found in the main body of this report and on the individual Quarterly Reports and associated Deliverables of the project, uploaded in PICS-NE.

  16. Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Pedro; Vale, Alberto [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ventura, Rodrigo [Institute for Systems and Robotics, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2015-07-01

    The Cask and Plug Remote Handling System (CPRHS) and the respective Cask Transfer System (CTS) are designed to transport activated components between the reactor and the hot cell buildings of ITER during maintenance operations. In nominal operation, the CPRHS/CTS shall operate autonomously under human supervision. However, in some unexpected situations, the automatic mode must be overridden and the vehicle must be remotely guided by a human operator due to the harsh conditions of the environment. The CPRHS/CTS is a rhombic-like vehicle with two independent steerable and drivable wheels along its longitudinal axis, giving it omni-directional capabilities. During manual guidance, the human operator has to deal with four degrees of freedom, namely the orientations and speeds of two wheels. This work proposes a Human Machine Interface (HMI) to manage the degrees of freedom and to remotely guide the CPRHS/CTS in ITER taking the most advantages of rhombic like capabilities. Previous work was done to drive each wheel independently, i.e., control the orientation and speed of each wheel independently. The results have shown that the proposed solution is inefficient. The attention of the human operator becomes focused in a single wheel. In addition, the proposed solution cannot assure that the commands accomplish the physical constrains of the vehicle, resulting in slippage or even in clashes. This work proposes a solution that consists in the control of the vehicle looking at the position of its center of mass and its heading in the world frame. The solution is implemented using a rotational disk to control the vehicle heading and a common analogue joystick to control the vector speed of the center of the mass of the vehicle. The number of degrees of freedom reduces to three, i.e., two angles (vehicle heading and the orientation of the vector speed) and a scalar (the magnitude of the speed vector). This is possible using a kinematic model based on the vehicle Instantaneous

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  18. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  19. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  20. Liprin-α2 promotes the presynaptic recruitment and turnover of RIM1/CASK to facilitate synaptic transmission.

    Science.gov (United States)

    Spangler, Samantha A; Schmitz, Sabine K; Kevenaar, Josta T; de Graaff, Esther; de Wit, Heidi; Demmers, Jeroen; Toonen, Ruud F; Hoogenraad, Casper C

    2013-06-10

    The presynaptic active zone mediates synaptic vesicle exocytosis, and modulation of its molecular composition is important for many types of synaptic plasticity. Here, we identify synaptic scaffold protein liprin-α2 as a key organizer in this process. We show that liprin-α2 levels were regulated by synaptic activity and the ubiquitin-proteasome system. Furthermore, liprin-α2 organized presynaptic ultrastructure and controlled synaptic output by regulating synaptic vesicle pool size. The presence of liprin-α2 at presynaptic sites did not depend on other active zone scaffolding proteins but was critical for recruitment of several components of the release machinery, including RIM1 and CASK. Fluorescence recovery after photobleaching showed that depletion of liprin-α2 resulted in reduced turnover of RIM1 and CASK at presynaptic terminals, suggesting that liprin-α2 promotes dynamic scaffolding for molecular complexes that facilitate synaptic vesicle release. Therefore, liprin-α2 plays an important role in maintaining active zone dynamics to modulate synaptic efficacy in response to changes in network activity.

  1. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  2. Evaluation of impact limiter performance during end-on and slapdown drop tests of a one-third scale model storage/transport cask system

    Energy Technology Data Exchange (ETDEWEB)

    Yoshimura, H.R.; Bronowski, D.R.; Uncapher, W.L.; Attaway, S.W.; Bateman, V.I.; Carne, T.G.; Gregory, D.L. (Sandia National Labs., Albuquerque, NM (USA)); Huerta, M. (Southwest Engineering Associates, El Paso, TX (USA))

    1990-12-01

    This report describes drop testing of a one-third scale model shipping cask system. Two casks were designed and fabricated by Transnuclear, Inc., to ship spent fuel from the former Nuclear Fuel Services West Valley reprocessing facility in New York to the Idaho National Engineering Laboratory for a long-term spent fuel dry storage demonstration project. As part of the NRC's regulatory certification process, one-third scale model tests were performed to obtain experimental data on impact limiter performance during impact testing. The objectives of the testing program were to (1) obtain deceleration and displacement information for the cask and impact limiter system, (2) obtain dynamic force-displacement data for the impact limiters, (3) verify the integrity of the impact limiter retention system, and (4) examine the crush behavior of the limiters. Two 30-ft (9-m) drop tests were conducted on a mass model of the cask body and scaled balsa and redwood-filled impact limiters. This report describes the results of both tests in terms of measured decelerations, posttest deformation measurements, and the general structural response of the system. 3 refs., 32 figs.

  3. Spectrum of pontocerebellar hypoplasia in 13 girls and boys with CASK mutations: confirmation of a recognizable phenotype and first description of a male mosaic patient

    Directory of Open Access Journals (Sweden)

    Burglen Lydie

    2012-03-01

    Full Text Available Abstract Background Pontocerebellar hypoplasia (PCH is a heterogeneous group of diseases characterized by lack of development and/or early neurodegeneration of cerebellum and brainstem. According to clinical features, seven subtypes of PCH have been described, PCH type 2 related to TSEN54 mutations being the most frequent. PCH is most often autosomal recessive though de novo anomalies in the X-linked gene CASK have recently been identified in patients, mostly females, presenting with intellectual disability, microcephaly and PCH (MICPCH. Methods Fourteen patients (12 females and two males; aged 16 months-14 years presenting with PCH at neuroimaging and with clinical characteristics unsuggestive of PCH1 or PCH2 were included. The CASK gene screening was performed using Array-CGH and sequencing. Clinical and neuroradiological features were collected. Results We observed a high frequency of patients with a CASK mutation (13/14. Ten patients (8 girls and 2 boys had intragenic mutations and three female patients had a Xp11.4 submicroscopic deletion including the CASK gene. All were de novo mutations. Phenotype was variable in severity but highly similar among the 11 girls and was characterized by psychomotor retardation, severe intellectual disability, progressive microcephaly, dystonia, mild dysmorphism, and scoliosis. Other signs were frequently associated, such as growth retardation, ophthalmologic anomalies (glaucoma, megalocornea and optic atrophy, deafness and epilepsy. As expected in an X-linked disease manifesting mainly in females, the boy hemizygous for a splice mutation had a very severe phenotype with nearly no development and refractory epilepsy. We described a mild phenotype in a boy with a mosaic truncating mutation. We found some degree of correlation between severity of the vermis hypoplasia and clinical phenotype. Conclusion This study describes a new series of PCH female patients with CASK inactivating mutations and confirms that

  4. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    Energy Technology Data Exchange (ETDEWEB)

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The

  5. Research on Shielding of High-capacity Cobalt Source Transport Cask%大容量钴源运输容器屏蔽研究

    Institute of Scientific and Technical Information of China (English)

    薛娜; 王炳衡; 毛亚蔚

    2015-01-01

    大容量钴源运输容器为运输工业用钴源而设计的专用设备。由于内容物放射性活度水平很高、衰变热很大,仅有少数国家具有设计能力,在国内的研制尚属首次。在对钴源运输容器的屏蔽设计研制过程中,突破之前的屏蔽设计技术束缚,采用MCAM程序与MCNP程序模拟计算钴源运输容器外的剂量率水平,并在设计过程中及时发现容器存在的设计缺陷,从而进行了设计改进,保证了容器满足国家标准要求的各项设计措施。目前这些设计措施已通过相关的试验验证。结果表明:针对大容量60 Co运输容器的关键技术制定的设计措施合理有效,充分保证了容器在经受国家标准中规定的正常运输条件和运输中事故条件下各项试验后容器屏蔽性能的完整性,确保钴源运输的安全。%High‐capacity cobalt source transport casks are used to transport 60 Co indus‐trial irradiators .The radioactive contents have special features of high‐activity and high residual heat ,so only a few countries have design capacity .This is the first design pro‐ject for the self‐reliant design of high‐capacity cobalt source transport casks .This paper was devoted to key technology in shielding design of these casks .The MCAM code and MCNP code were used for the calculation of the dose rate level outside the cask and the design improvement was applied in the cask to meet the requirements in national stand‐ard .A series of test proved the casks have ability to transport high‐activity sealed sources safely .Calculation results in design are in well concordance with survey results . It demonstrates the rationality and reliability of the methods used in this shielding design .The patent for the design of high‐capacity cobalt source transport casks was obtained .Through the design for cobalt source transport casks ,a good foundation is laid for the self‐reliant design of spent fuel

  6. A methodology to quantify the release of spent nuclear fuel from dry casks during security-related scenarios.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Luna, Robert Earl

    2013-11-01

    Assessing the risk to the public and the environment from a release of radioactive material produced by accidental or purposeful forces/environments is an important aspect of the regulatory process in many facets of the nuclear industry. In particular, the transport and storage of radioactive materials is of particular concern to the public, especially with regard to potential sabotage acts that might be undertaken by terror groups to cause injuries, panic, and/or economic consequences to a nation. For many such postulated attacks, no breach in the robust cask or storage module containment is expected to occur. However, there exists evidence that some hypothetical attack modes can penetrate and cause a release of radioactive material. This report is intended as an unclassified overview of the methodology for release estimation as well as a guide to useful resource data from unclassified sources and relevant analysis methods for the estimation process.

  7. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  8. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  9. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.

    2014-06-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  10. Determination of uncertainties in the calculation of dose rates at transport and storage casks; Unsicherheiten bei der Berechnung von Dosisleistungen an Transport- und Lagerbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Schloemer, Luc Laurent Alexander

    2014-12-17

    The compliance with the dose rate limits for transport and storage casks (TLB) for spent nuclear fuel from pressurised water reactors can be proved by calculation. This includes the determination of the radioactive sources and the shielding-capability of the cask. In this thesis the entire computational chain, which extends from the determination of the source terms to the final Monte-Carlo-transport-calculation is analysed and the arising uncertainties are quantified not only by benchmarks but also by variational calculi. The background of these analyses is that the comparison with measured dose rates at different TLBs shows an overestimation by the values calculated. Regarding the studies performed, the overestimation can be mainly explained by the detector characteristics for the measurement of the neutron dose rate and additionally in case of the gamma dose rates by the energy group structure, which the calculation is based on. It turns out that the consideration of the uncertainties occurring along the computational chain can lead to even greater overestimation. Concerning the dose rate calculation at cask loadings with spent uranium fuel assemblies an uncertainty of (({sup +21}{sub -28}) ±2) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are estimated. For mixed-loadings with spent uranium and MOX fuel assemblies an uncertainty of ({sup +24±3}{sub -27±2}) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are quantified. The results show that the computational chain has not to be modified, because the calculations performed lead to conservative dose rate predictions, even if high uncertainties at neutron dose rate measurements arise. Thus at first the uncertainties of the neutron dose rate measurement have to be decreased to enable a reduction of the overestimation of the calculated dose rate afterwards. In the present thesis

  11. Study of the temperature distribution on welded thin plates of duplex steel to be used for the external clad of a cask for transportation of radiopharmaceuticals products

    Energy Technology Data Exchange (ETDEWEB)

    Betini, Evandro G.; Ceoni, Francisco C.; Mucsi, Cristiano S.; Politano, Rodolfo; Rossi, Jesualdo L., E-mail: egbetini@ipen.br, E-mail: fceoni@hotmail.com, E-mail: csmucsi@ipen.br, E-mail: politano@ipen.br, E-mail: jelrossi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Orlando, Marcos T.D., E-mail: mtdorlando@gmail.com [Universidade Federal do Espirito Santo (CCE/DFIS/UFES), Vitoria, ES (Brazil). Centro de Ciencias Exatas. Departamento de Fisica

    2015-07-01

    The clad material for a proprietary transport device for radiopharmaceutical products is the main focus of the present work. The production of {sup 99}Mo-{sup 99m}Tc transport cask requires a receptacle or cask where the UNS S32304 duplex steel sheet has shown that it meets high demands as the required mechanical strength and the spread of impact or shock waves mitigation. This work reports the experimental efforts in recording the thermal distribution on autogenous thin plates of UNS S32304 steel during welding. The UNS S32304 duplex steel is the most probable candidate for the external clad of the containment package for the transport of radioactive substances so it is highly relevant the understanding of all its physical parameters and its behavior under the thermal cycle imposed by a welding process. For the welding of the UNS S32304 autogenous plates the GTAW (gas tungsten arc welding) process was used with a pure argon arc protection atmosphere in order to simulate a butt joint weld on a thin duplex steel plate without filler metal. The thermal cycles were recorded by means of K-type thermocouples embedded by electrical spot welding near the weld region and connected to a multi-channel data acquisition system. The obtained results validate the reliability of the experimental apparatus for the future complete analysis of the welding experiment and further comparison to numerical analysis. (author)

  12. Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

    2012-07-06

    The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that

  13. Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

    2012-07-06

    The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that

  14. GY-20和GY-40型大容量钴-60运输容器关键技术研究%Key Technology Studies of GY-20 and GY-40 High-Capacity Cobalt-60 Transport Casks

    Institute of Scientific and Technical Information of China (English)

    刘慧芳; 张鑫

    2012-01-01

    GY -20和GY- 40型大容量钴-60运输容器是为运输工业用钴-60成品源和钴-60棒束而设计的专用设备.由于内容物放射性活度水平高、衰变热大,仅有加拿大、英国、俄罗斯等少数国家具有设计能力.本文综合考虑容器结构、热工、力学、屏蔽等方面的要求,对容器设计和制造过程中的关键技术以及解决方案进行了分析研究.试验验证结果表明,容器的结构设计、包铅边界设计准则的确定和制造过程的质量控制措施合理、有效,能保证容器在各种工况下的屏蔽完整性,容器具有安全运输大容量钴-60源项的能力,其设计满足相关标准和规范要求,可为其他B型货包的设计提供参考.%GY-20 and GY-40 high-capacity cobalt-60 transport,casks are used to transport cobalt-60 industrial irradiators and cobalt-60 bundles. The radioactive contents have special features of high-activity and high residual heat, so only a few countries such as Canada, England and Russia have design capacity. The key technologies and corresponding solutions were studied for the design and manufacture of the cask taking into account the structural, thermal, mechanics and shield requests. A series of tests prove that the cask structure design, design criteria for lead coating structure and quality control measurements are reasonable and effective, and the cask shield integrity can be ensured for all conditions. The casks have ability to transport high-activity sealed sources safely, and the design of cask satisfies the requirement of design code and standard. It can provide reference for other B type package.

  15. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  16. Conceptual design of a high-integrity impact limiter for use in shipment of dual-purpose spent-fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (United States)); Haelsig, R.T.; Hansen, L.J. (Hansen Haelsig Associates, Bellevue, Washington (USA))

    1991-09-01

    A conceptual design for a high-integrity impact limiting system to protect dry metallic spent fuel storage casks during rail transport is proposed. The system is intended to limit the deceleration of the cask during severe rail accidents through three layers of energy-absorbing polyurethane foam material. The crush strengths of the foam is chosen such that the lowest crush strength foam forms the most exterior layer, with the crush strengths increasingly progressively in the two inner layers. The design basis for the external layer of foam is the hypothetical 30-foot free drop impact event prescribed in 10 CFR 71, with a peak steady deceleration limit of about 75 g. The two interior layers absorb up to five times the impact energy of the 30-foot free drop while limiting the decelerations to first 125 g and then to 175 g. The former is felt to be a nominal fuel rod failure threshold, while the latter is at or near the failure level for bolted closure assemblies. These deceleration targets, if met, provide a means for substantially reducing the risk of radioactive material transport. The conceptual design incorporates features for maintaining the integrity of the impact limiter attachment system during severe accidents and enhancing heat dissipation through the impact limiter for short-cooled fuel, through the use of radial aluminum fins. An alternative impact-limiting material -- aluminum honeycomb -- is included in the economic assessment. Both the polyurethane foam and aluminum honeycomb designs appear to meet a cost target of $1.0M, with the polyurethane foam limiter cost estimated at somewhat less than $400K and the aluminum honeycomb cost at somewhat less than $700K. 28 refs., 17 figs., 5 tabs.

  17. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis; Untersuchung des Verhaltens stossdaempfender Bauteile von Transportbehaeltern fuer radioaktive Stoffe in Bauartpruefung und Risikoanalyse

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  18. Decree no. 2001-1199 of the 10 december 2001 publishing the resolution MSC. 88 (71) notifying adoption of the international compilation of safety rules for the spent nuclear fuels, plutonium and high level radioactive wastes transport in casks on ships (compilation INF) (annexes), adopted at London the 27 may 1999; Decret no. 2001-1199 du 10 decembre 2001 portant publication de la resolution MSC.88 (71) portant adoption du recueil international de regles de securite pour le transport de combustible nucleaire irradie, de plutonium et de dechets hautement radioactifs en colis a bord de navires (recueil INF) (ensemble une annexe), adoptee a Londres le 27 mai 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This legislative text concerns the safety rules of spent nuclear fuels, plutonium and high level radioactive wastes transport, in casks on ships. Rules, fire prevention, temperature control of casks, electric supply, radioprotection, management and emergency plans are detailed. (A.L.B.)

  19. Loading 076 assemblies in two IV-04 transport casks for transport to the U.S. Savannah River Site (SC); Trasferimento di 72 elementi irraggiati MTR dalla piscina dell`impianto EUREX a due contenitori IU-04 per il trasporto al Savannah River Site-Department of Energy (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Gili, Michele [ENEA, Centro Ricerche Saluggia, Vercelli (Italy). Dipt. Energia

    1997-09-01

    The National Agency for New Technologies and the Environments has signed with the US Department of Energy a contract for the transfer of 150 irradiated MTR fuel assemblies stored in the EUREX plant pool at The National Agency for New Technologies and the Environments Research Centre of Saluggia. The first scheduled transport has been made in february 1997 and has involved the successful loading of 76 assemblies in two IU-04 (Pegase) transport casks. The loaded casks have been shipped to the U.S. Savannah River Site (SC).

  20. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis; Untersuchung des Verhaltens stossdaempfender Bauteile von Transportbehaeltern fuer radioaktive Stoffe in Bauartpruefung und Risikoanalyse

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  1. Perfil físico-químico de aguardente durante envelhecimento em tonéis de carvalho Chemical profile of aguardente - Brazilian sugar cane alcoholic drink - aged in oak casks

    Directory of Open Access Journals (Sweden)

    Mariana Branco de Miranda

    2008-12-01

    Full Text Available Avaliou-se por um período de 390 dias o perfil da composição química da aguardente sob envelhecimento em tonéis de carvalho de 20 L. O envelhecimento da aguardente em tonéis de madeira melhora a qualidade sensorial do destilado. As aguardentes envelhecidas foram analisadas aos 0, 76, 147, 228, 314 e 390 dias de armazenamento quanto às concentrações de etanol, acidez volátil, ésteres, aldeídos, furfural, álcoois superiores (n-propílico, isobutílico e isoamílicos, metanol, cobre, extrato seco, taninos e cor. Após os 390 dias de armazenamento, a aguardente apresentou maiores concentrações de acidez volátil, ésteres, aldeídos, furfural, álcoois superiores, congêneres, extrato seco e tanino. Sua coloração tornou-se amarelada. As concentrações de etanol e de metanol não se alteraram, e o teor de cobre apresentou ligeiro declínio. O envelhecimento da aguardente por 390 dias em tonéis de carvalho alterou a sua composição química, porém ela se manteve dentro de todos os padrões de qualidade estabelecidos pela legislação nacional em vigor.The chemical composition of aguardente - Brazilian sugar cane alcoholic drink - under aging during in 20 L oak casks was evaluated for 390 days. Aging sugar cane aguardente in wood casks improves the sensorial quality of the distillate. The concentrations of ethanol, volatile acidity, esters, aldehydes, furfural, higher alcohols (n-propylic, isobutylic and isoamylics, methanol, copper, dry extract, tannins, and color of the aged sugar cane aguardente were analysed at 0, 76, 147, 228, 314, and 390 days of storage. After 390 days of aging the sugar cane aguardente presented higher concentrations of volatile acidity, esters, aldehydes, furfural, higher alcohols, congeners, dry extract, and tannin. Its color became golden. The concentrations of ethanol and methanol did not change and the copper content decreased slightly. The aging of the sugar cane aguardente in oak casks for 390 days

  2. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage; Estudo de um casco nacional e sua instalacao para armazenagem seca de combustivel nuclear queimado gerado em reatores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio

    2009-07-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  3. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  4. The Knee Clinical Assessment Study – CAS(K. A prospective study of knee pain and knee osteoarthritis in the general population: baseline recruitment and retention at 18 months

    Directory of Open Access Journals (Sweden)

    Duncan Rachel

    2006-03-01

    Full Text Available Abstract Background Selective non-participation at baseline (due to non-response and non-consent and loss to follow-up are important concerns for longitudinal observational research. We investigated these matters in the context of baseline recruitment and retention at 18 months of participants for a prospective observational cohort study of knee pain and knee osteoarthritis in the general population. Methods Participants were recruited to the Knee Clinical Assessment Study – CAS(K – by a multi-stage process involving response to two postal questionnaires, consent to further contact and medical record review (optional, and attendance at a research clinic. Follow-up at 18-months was by postal questionnaire. The characteristics of responders/consenters were described for each stage in the recruitment process to identify patterns of selective non-participation and loss to follow-up. The external validity of findings from the clinic attenders was tested by comparing the distribution of WOMAC scores and the association between physical function and obesity with the same parameters measured directly in the target population as whole. Results 3106 adults aged 50 years and over reporting knee pain in the previous 12 months were identified from the first baseline questionnaire. Of these, 819 consented to further contact, responded to the second questionnaire, and attended the research clinics. 776 were successfully followed up at 18 months. There was evidence of selective non-participation during recruitment (aged 80 years and over, lower socioeconomic group, currently in employment, experiencing anxiety or depression, brief episode of knee pain within the previous year. This did not cause significant bias in either the distribution of WOMAC scores or the association between physical function and obesity. Conclusion Despite recruiting a minority of the target population to the research clinics and some evidence of selective non-participation, this

  5. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  6. The Fusion of Horror and Aesthetics:Gothic Narration in The Cask of Amontillado%恐怖与美学的融合:《阿芒提拉多酒桶》的哥特式叙事

    Institute of Scientific and Technical Information of China (English)

    孙峰

    2014-01-01

    As an important way of narration, Gothic style is vividly and incisively applied in many of Edgar Allan Poe’s short stories, showing quite a lot of horrible and scary scenes for the readers. While reaching to an extreme, the horror may turn to its opposite side and presents the story with a strange beauty of literature, which is bestowed with unconventional aesthetic characteristics. As one of Edgar Allan Poe’s masterpieces, the short story The Cask of Amontillado deals with the theme of revenge and death. It contains only a few characters and its plot is not too complex. However, the story has a rich atmosphere of horror, which conveys exactly the Gothic style and achieves an unusual artistic effect.%作为一种重要的叙事方式,哥特手法在爱伦·坡的众多短篇小说中得到了淋漓尽致的运用,为读者展现了一幕幕令人惊悚恐惧的场景。这种恐怖达到一个极致,便向它的相反面转化,使作为载体的故事情节呈现了一种异样的文学之美,从而被赋予了反传统的美学艺术特征。短篇小说《阿芒提拉多酒桶》作为爱伦·坡的短篇代表作之一,以复仇和死亡为主题,人物不多,情节也不太复杂,但却有着浓郁的恐怖气氛,恰到好处地传达出这种创作特征,实现了异乎寻常的艺术效果。

  7. Aspectos da composição química e aceitação sensorial da aguardente de cana-de-açúcar envelhecida em tonéis de diferentes madeiras Aspects of the chemical composition and sensorial acceptance of sugar cane spirit aged in casks of different types of woods

    Directory of Open Access Journals (Sweden)

    André Ricardo Alcarde

    2010-05-01

    sugar cane spirit aged for 3 years in casks of different types of wood (peanut wood, araruva or striped wood, red cabreuva, oak, cherrywood, Brazilian gold wood, purple tabebuia, cariniana legalis, and pear tree. The simple alcoholic distillate which originated the sugar cane spirit was produced at the Distillery of ESALQ/USP. After aging, the sugar cane spirits were analyzed in terms of ethanol concentrations o, volatile acidity, furfural, aldehydes, esters, higher alcohols, methanol, copper, total phenolic compounds, color, and sensorial acceptance. Regardless the type of wood the casks were made of, the aged sugar cane spirits became darker and presented higher concentrations of volatile acidity, furfural, esters, higher alcohols, congeners, and total phenolic compounds than the simple alcoholic distillate. On the other hand, the aged sugar cane spirits presented lower concentrations of aldehydes, methanol, and copper than the simple alcoholic distillate. The statistical analysis, considering the global physicochemical composition of the sugar cane spirits aged in the casks made of different types of wood, showed similarities among the sugar cane spirits aged in the casks of peanut wood, araruva or striped wood, and cariniana legalis. It also indicates similarities among the sugar cane spirits aged in the casks of red cabreuva and pear tree and among the sugar cane spirits aged in the casks of oak, cherrywood, Brazilian gold wood, and purple tabebuia. The sugar cane spirits aged in the casks of the different types of wood were in accordance with the composition and quality standards established by the Brazilian laws. The sugar cane spirit aged in oak presented the best sensorial acceptance. Among the Brazilian woods, purple tabebuia, peanut wood, red cabreuva, cherrywood and pear tree were those that produced sugar cane spirits with better sensorial qualities.

  8. Finite-element-method-based assessment on the dropping accident of an high temperature gas cooled reactor fuel cask%基于有限元方法的高温气冷堆燃料贮存罐跌落事故评价

    Institute of Scientific and Technical Information of China (English)

    聂君锋; 张海泉; 李红克; 王鑫; 张征明

    2013-01-01

    通过将燃料元件等效为流体,本文采用耦合Eulerian-Lagrangian(CEL)方法研究了高温气冷堆燃料元件贮存罐的跌落事故.该方法能够描述燃料元件在跌落过程中的流动性和惯性效应,以及燃料元件对贮存罐所产生的侧向液动压力.与等效质量法进行了对比,结果表明:在跌落冲击过程中,等效质量法计算得到的冲击力更大、跌落接触时间更短,而CEL方法则能体现罐体的径向膨胀.因此,CEL方法能够模拟燃料元件的惯性效应以及流动效应,而等效质量法则能充分考虑冲击力的作用,结构设计中可以结合2种方法的计算结果,给出更为合理的设计方案.%Accidental dropping of an HTGR (high temperature gas cooled reactor) fuel cask was analyzed using the Euler-Lagrange (CEL) method with the fuel element modeled as the fluid.The method can describe the flow and inertial effects of the fuel elements during the fall and the lateral fluid dynamic pressure generated by the fuel element on the fuel cask.The results give a larger impact force than the equivalent mass method with a shorter drop time.The CEL method can also predict the radial expansion of the cask.Therefore,the CEL method is able to simulate the inertial effects and the liquidity effects of the fuel element,while the equivalent mass method analyzes only the impact.The results of the two methods can be combined in the structural design to give a more reasonable design.

  9. Report on task assignment No. 3 for the Waste Package Project; Parts A & B, ASME pressure vessel codes review for waste package application; Part C, Library search for reliability/failure rates data on low temperature low pressure piping, containers, and casks with long design lives

    Energy Technology Data Exchange (ETDEWEB)

    Trabia, M.B.; Kiley, M.; Cardle, J.; Joseph, M.

    1991-07-01

    The Waste Package Project Research Team, at UNLV, has four general required tasks. Task one is the management, quality assurance, and overview of the research that is performed under the cooperative agreement. Task two is the structural analysis of spent fuel and high level waste. Task three is an American Society of Mechanical Engineers (ASME) Pressure Vessel Code review for waste package application. Finally, task four is waste package labeling. This report includes preliminary information about task three (ASME Pressure Vessel Code review for Waste package Application). The first objective is to compile a list of the ASME Pressure Vessel Code that can be applied to waste package containers design and manufacturing processes. The second objective is to explore the use of these applicable codes to the preliminary waste package container designs. The final objective is to perform a library search for reliability and/or failure rates data on low pressure, low temperature, containers and casks with long design lives.

  10. Packaging design criteria for the MCO cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-01-30

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the K Basins to a Canister Storage Building in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks.

  11. Nondestructive Examination Guidance for Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lareau, John P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhuge, Jing Wei [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Moran, Traci L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure-Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific components.

  12. Discussion on the Gothic features in A Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    刘亚平

    2011-01-01

    Edgar Allan Poe is a writer who establishes a new school in the American history of literature.He firstly and consciously takes the short story as an independent literature style,advances his own composing theory of short stories and applies it to his com

  13. Project W-443 cask/transporation project management plan

    Energy Technology Data Exchange (ETDEWEB)

    Byrd, L.C., Westinghouse Hanford

    1996-07-02

    This document has been prepared and is being released for Project W-443 participants to use in the performance of project activities. This PMP establishes the organizational responsibilities and baseline controls to be used to manage Spent Nuclear Fuel Subproject W-443.

  14. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

    ... two main forms: microcephaly with pontine and cerebellar hypoplasia (MICPCH), and X-linked intellectual disability (XL-ID) ... progressive microcephaly). Individuals with this condition have underdevelopment (hypoplasia) of areas of the brain called the cerebellum ...

  15. Probabilistic Multi-Hazard Assessment of Dry Cask Structures

    Energy Technology Data Exchange (ETDEWEB)

    Bencturk, Bora [Univ. of Houston, TX (United States); Padgett, Jamie [Rice Univ., Houston, TX (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States).

    2017-01-10

    systems the concrete shall not only provide shielding but insures stability of the upright canister, facilitates anchoring, allows ventilation, and provides physical protection against theft, severe weather and natural (seismic) as well as man-made events (blast incidences). Given the need to remain functional for 40 years or even longer in case of interim storage, the concrete outerpack and the internal canister components need to be evaluated with regard to their long-term ability to perform their intended design functions. Just as evidenced by deteriorating concrete bridges, there are reported visible degradation mechanisms of dry storage systems especially when high corrosive environments are considered in maritime locations. The degradation of reinforced concrete is caused by multiple physical and chemical mechanisms, which may be summarized under the heading of environmental aging. The underlying hygro-thermal transport processes are accelerated by irradiation effects, hence creep and shrinkage need to include the effect of chloride penetration, alkali aggregate reaction as well as corrosion of the reinforcing steel. In light of the above, the two main objectives of this project are to (1) develop a probabilistic multi-hazard assessment framework, and (2) through experimental and numerical research perform a comprehensive assessment under combined earthquake loads and aging induced deterioration, which will also provide data for the development and validation of the probabilistic framework.

  16. Property enhancement of cast iron used for nuclear casks

    Science.gov (United States)

    Behera, R. K.; Mahto, B. P.; Dubey, J. S.; Mishra, S. C.; Sen, S.

    2016-01-01

    Ductile iron (DI) is a preferred material for use in various structural, automotive, and engineering fields because of its excellent combination of strength, toughness, and ductility. In the current investigation, we elucidate the relationship between the morphological and mechanical properties of DI intended for use in safety applications in the nuclear industry. DI specimens with various alloying elements were subjected to annealing and austempering heat treatment processes. A faster cooling rate appeared to increase the nodule count in austempered specimens, compensating for their nodularity value and subsequently decreasing their ductility and impact strength. The ductility and impact energy values of annealed specimens increased with increasing ferrite area fraction and nodularity, whereas an increase in the amounts of Ni and Cr resulted in an increase of hardness via solid solution strengthening. Austempered specimens were observed to be stronger than annealed specimens and failed in a somewhat brittle manner characterized by a river pattern, whereas the ductile failure mode was characterized by the presence of dimples.

  17. Property enhancement of cast iron used for nuclear casks

    Institute of Scientific and Technical Information of China (English)

    RK Behera; BP Mahto; JS Dubey; SC Mishra; S Sen

    2016-01-01

    Ductile iron (DI) is a preferred material for use in various structural, automotive, and engineering fields because of its excellent combination of strength, toughness, and ductility. In the current investigation, we elucidate the relationship between the morphological and mechanical properties of DI intended for use in safety applications in the nuclear industry. DI specimens with various alloying elements were subjected to annealing and austempering heat treatment processes. A faster cooling rate appeared to increase the nodule count in austempered specimens, compensating for their nodularity value and subsequently decreasing their ductility and impact strength. The ductility and impact energy values of annealed specimens increased with increasing ferrite area fraction and nodularity, whereas an increase in the amounts of Ni and Cr resulted in an increase of hardness via solid solution strengthening. Austempered specimens were observed to be stronger than an-nealed specimens and failed in a somewhat brittle manner characterized by a river pattern, whereas the ductile failure mode was character-ized by the presence of dimples.

  18. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... National Technology Transfer and Advancement Act of 1995 (Pub. L. 104-113) requires that Federal agencies... spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation... explains why the rule would be inappropriate, including challenges to the rule's underlying premise...

  19. Validation Experiments for Spent- Fuel Dry-Cask In-Basket Convection

    Energy Technology Data Exchange (ETDEWEB)

    Smith, barton [Utah State Univ., Logan, UT (United States)

    2016-08-16

    This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

  20. Validation Experiments for Spent-Fuel Dry-Cask In-Basket Convection

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Barton L. [Utah State Univ., Logan, UT (United States). Dept. of Mechanical and Aerospace Engineering

    2016-08-16

    This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

  1. Comparison of cask and drywell storage concepts for a monitored retrievable storage/interim storage system

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, D.E.

    1982-12-01

    The Department of Energy, through its Richland Operations Office is evaluating the feasibility, timing, and cost of providing a federal capability for storing the spent fuel, high-level wastes, and transuranic wastes that DOE may be obligated by law to manage until permanent waste disposal facilities are available. Three concepts utilizing a monitored retrievable storage/interim storage (MRS/IS) facility have been developed and analyzed. The first concept, co-location with a reprocessing plant, has been developed by staff of Allied General Nuclear Services. the second concept, a stand-alone facility, has been developed by staff of the General Atomic Company. The third concept, co-location with a deep geologic repository, has been developed by the Pacific Northwest Laboratory with the assistance of the Westinghouse Hanford Company and Kaiser Engineers. The objectives of this study are: to develop preconceptual designs for MRS/IS facilities: to examine various issues such as transportation of wastes, licensing of the facilities, and environmental concerns associated with operation of such facilities; and to estimate the life-cycle costs of the facilities when operated in response to a set of scenarios that define the quantities and types of waste requiring storage in specific time periods, generally spanning the years 1989 to 2037. Three scenarios are examined to develop estimates of life-cycle costs for the MRS/IS facilities. In the first scenario, the reprocessing plant is placed in service in 1989 and HLW canisters are stored until a repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at intervals as needed to meet the demand. In the second scenario, the reprocessing plants are delayed in starting operations by 10 years, but the repositories open on schedule. In the third scenario, the repositories are delayed 10 years, but the reprocessing plants open on schedule.

  2. 75 FR 41404 - List of Approved Spent Fuel Storage Casks: NUHOMS®

    Science.gov (United States)

    2010-07-16

    ... average size of the boron carbide ] particles in the finished product is approximately 50 microns after....3 of the SAR, are not precisely quantified in that it requires that ``the average size of the boron carbide particles in the finished product is approximately 50 microns after rolling.'' Use of...

  3. 75 FR 41369 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD Revision 1; Withdrawal

    Science.gov (United States)

    2010-07-16

    ... boron carbide particles in the finished product is approximately 50 microns after rolling.'' Use of... average size of the boron carbide particles in the finished product is approximately 50 microns after... FR 24786), is withdrawn. FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Federal...

  4. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR- RELATED GREATER THAN CLASS C WASTE 1. The... the NUHOMS HD Horizontal Modular Storage System for Irradiated Nuclear Fuel. ] Docket Number: 72-1030... From the Federal Register Online via the Government Publishing Office NUCLEAR...

  5. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... Standardized NUHOMS® Horizontal Modular Storage System for Irradiated Nuclear Fuel. Docket Number:...

  6. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-07

    ... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR- RELATED GREATER THAN CLASS C... Safety Analysis Report for the NUHOMS HD Horizontal Modular Storage System for Irradiated Nuclear Fuel...; #0; #0;#0;Federal Register / Vol. 75, No. 88 / Friday, May 7, 2010 / Proposed Rules#0;#0; ]...

  7. 76 FR 70374 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... ID NRC- 2011-0008. Address questions about NRC dockets to Carol Gallagher, telephone: (301) 492-3668, email: Carol.Gallagher@nrc.gov . Mail comments to: Secretary, U.S. Nuclear Regulatory Commission..., 2011. For the Nuclear Regulatory Commission. Michael F. Weber, Acting Executive Director for...

  8. The Structural Design and Analysis of Pallet in ITER Transfer Cask for Remote Handling Operations

    Science.gov (United States)

    Zhou, Zibo; Yao, Damao; Cao, Lei; Li, Ge

    2009-06-01

    Necessary adjustment ranges and accuracies of the pallet for ITER are presented. Detailed structural designs and structural finite element analyses for pallet components are made to determine whether the results satisfy the requirements of the pallet structure to be used in ITER.

  9. The Structural Design and Analysis of Pallet in ITER Transfer Cask for Remote Handling Operations

    Institute of Scientific and Technical Information of China (English)

    ZHOU Zibo; YAO Damao; CAO Lei; LI Ge

    2009-01-01

    Necessary adjustment ranges and accuracies of the pallet for ITER are presented. Detailed structural designs and structural finite element analyses for pallet components are made to determine whether the results satisfy the requirements of the pallet structure to be used in ITER.

  10. 76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... Federal Regulations is sold by the Superintendent of Documents. #0;Prices of new books are listed in the... Combustion Engineering 16x16 class fuel assemblies as authorized contents; reduce the minimum off-normal... Engineering 16x16 class fuel assemblies as authorized contents; reduce the minimum off-normal...

  11. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Science.gov (United States)

    2010-01-01

    ... Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... adversely affected structures, systems, and components important to safety....

  12. 75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-06

    ... include pressurized water reactor fuel assemblies with control components, reduce the minimum initial..., for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective... sites of civilian nuclear power reactors without, to the maximum extent practicable, the need...

  13. Natural ventilation of a generic cask under a transport hood - CFD and analytical modelling

    Energy Technology Data Exchange (ETDEWEB)

    Powell, D.; Davies, G.; Tso, C.F. [Arup, London (United Kingdom)

    2004-07-01

    In comparison with finite element simulation for structural and thermal behaviour, the use of computational fluid dynamics technique (hereafter CFD) to analyse, predict and design air and heat flow in package design is relatively novel. Arup has been using CFD techniques to investigate fluid and heat flow, and to use it as a tool to design fluid and heat flow across a broad spectrum of industries for over fifteen years. In order demonstrate the power of the technique and its benefits, the airflow and heat flow characteristics around a transport package during transit under a transport hood has been evaluated using the CFD technique. This paper presents the scenario, the model, the analysis technique and the results of this analysis. Comparison with test results is probably the best way to validate a CFD analysis. In the absence of test results, the analysis was verified by comparison with hand calculation solutions. The scenario as it stands is too complex and hand calculation solution cannot describe the scenario sufficiently. However, hand calculation solutions could be derived for simplified version of the scenario against which CFD analysis of the simplified scenario can be compared. The second half of this paper describes the verification out.

  14. A proposal for an international brittle fracture acceptance criterion for nuclear material transport cask applications

    Energy Technology Data Exchange (ETDEWEB)

    Sorenson, K.B.; Salzbrenner, R.J.; Nickell, R.E.

    1989-01-01

    This paper presents a fundamental basis for a brittle fracture acceptance criterion, examine several existing criteria and propose examples for consideration as international brittle fracture acceptance criteria. The proposed criteria are intended to stimulate discussion in order to advance the development of a consensus approach. 8 refs., 1 fig., 1 tab.

  15. 78 FR 37927 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-06-25

    ... Federal Regulations is sold by the Superintendent of Documents. #0;Prices of new books are listed in the... revises authorized contents to include: pressurized water reactor (PWR) damaged fuel contained in damaged.... Tanious, Office of Federal and State Materials and Environmental Management Programs, U.S....

  16. Modelling of pool fire environments using experimental results of a two-hour test of a railcar/cask system

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, J.E.; Klein, D.E.; Pope, R.B.; Yoshimura, H.R.

    1980-01-01

    It was demonstrated that time and spatial variations in the local source temperatures, the radiant shielding of intervening structure and the effects of wind can significantly affect the amount of heat input to a large package in a simulated accidental fire. The pool fire provided a significantly non-uniform heat source to the package. Despite these effects, however, the amount of heat input to the package was generally equivalent to that which would be received from a regulatory 800/sup 0/C uniform thermal source. 7 firegures.

  17. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... Reactor (BWR) fuel with high initial enrichment (up to 4.8 weight percent uranium-235 planer average...) The ability to store and transport BWR fuel with high initial enrichment (up to 4.8 weight percent... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR...

  18. Theory from the cask harmonious team-building%从木桶理论看和谐团队建设

    Institute of Scientific and Technical Information of China (English)

    阚卫华; 宋波; 崔坤家

    2007-01-01

    现代企业的团队建设与木桶理论有着异曲同工之妙.加强企业的团队建设,应注重加固"长板"以发挥优势;弥补"短板"以提升能力;进行团队协作,构建文化支撑.

  19. Actinide Partitioning-Transmutation Program Final Report. V. Preconceptual designs and costs of partitioning facilities and shipping casks (appendix 3)

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    This Appendix contains cost estimate documents for the Fuels Reprocessing Plant Waste Treatment Facility. Plant costs are summarized by Code of Accounts and by Process Function. Costs contribution to each account are detailed. Process equipment costs are detailed for each Waste Treatment Process. Service utility costs are also summarized and detailed.

  20. 10 CFR 72.103 - Geological and seismological characteristics for applications for dry cask modes of storage on or...

    Science.gov (United States)

    2010-01-01

    ... conditions that must be evaluated include soil and rock stability, liquefaction potential, and natural and... unstable geological characteristics, soil stability problems, or potential for vibratory ground motion at... be used. (c) Sites other than bedrock sites must be evaluated for their liquefaction potential...

  1. 76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

    Science.gov (United States)

    2011-03-09

    ... Federal Regulations is sold by the Superintendent of Documents. #0;Prices of new books are listed in the...; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16...

  2. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-23

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  3. Gender difference in symptomatic radiographic knee osteoarthritis in the Knee Clinical Assessment – CAS(K: A prospective study in the general population

    Directory of Open Access Journals (Sweden)

    Duncan Rachel C

    2008-06-01

    Full Text Available Abstract Background A recent study of adults aged ≥50 years reporting knee pain found an excess of radiographic knee osteoarthritis (knee ROA in symptomatic males compared to females. This was independent of age, BMI and other clinical signs and symptoms. Since this finding contradicts many previous studies, our objective was to explore four possible explanations for this gender difference: X-ray views, selection, occupation and non-articular conditions. Methods A community-based prospective study. 819 adults aged ≥50 years reporting knee pain in the previous 12 months were recruited by postal questionnaires to a research clinic involving plain radiography (weight-bearing posteroanterior semiflexed, supine skyline and lateral views, clinical interview and physical examination. Any knee ROA, ROA severity, tibiofemoral joint osteoarthritis (TJOA and patellofemoral joint osteoarthritis (PJOA were defined using all three radiographic views. Occupational class was derived from current or last job title. Proportions of each gender with symptomatic knee ROA were expressed as percentages, stratified by age; differences between genders were expressed as percentage differences with 95% confidence intervals. Results 745 symptomatic participants were eligible and had complete X-ray data. Males had a higher occurrence (77% of any knee ROA than females (61%. In 50–64 year olds, the excess in men was mild knee OA (particularly PJOA; in ≥65 year olds, the excess was both mild and moderate/severe knee OA (particularly combined TJOA/PJOA. This male excess persisted when using the posteroanterior view only (64% vs. 52%. The lowest level of participation in the clinic was symptomatic females aged 65+. Within each occupational class there were more males with symptomatic knee ROA than females. In those aged 50–64 years, non-articular conditions were equally common in both genders although, in those aged 65+, they occurred more frequently in symptomatic females (41% than males (31%. Conclusion The excess of knee ROA among symptomatic males in this study seems unlikely to be attributable to the use of comprehensive X-ray views. Although prior occupational exposures and the presence of non-articular conditions cannot be fully excluded, selective non-participation bias seems the most likely explanation. This has implications for future study design.

  4. The Role of the Neurofibromin-Syndecan-Cask Complex in the Regulation of Synlaptic RAS-MAPK Signaling and Denoritic Spine Plasticity

    Science.gov (United States)

    2007-02-01

    dendritic spine remodeling using both DG- CA3 explants and CA1/3 slice cultures. With the DG explants, we showed that Nf1+/- also displayed some subtle...or Akt, or treated with upstream activator, BDNF ( Kumar et al., 2005). These novel exciting observations should have important implications for

  5. The Knee Clinical Assessment Study – CAS(K). A prospective study of knee pain and knee osteoarthritis in the general population

    OpenAIRE

    Hay Elaine; Duncan Rachel; Wilkie Ross; Myers Helen; Dziedzic Krysia; Wood Laurence; Handy June; Thomas Elaine; Peat George; Hill Jonathan; Croft Peter

    2004-01-01

    Abstract Background Knee pain affects an estimated 25% of the adult population aged 50 years and over. Osteoarthritis is the most common diagnosis made in older adults consulting with knee pain in primary care. However, the relationship between this diagnosis and both the current disease-based definition of osteoarthritis and the regional pain syndrome of knee pain and disability is unclear. Expert consensus, based on current evidence, views the disease and the syndrome as distinct entities b...

  6. The Knee Clinical Assessment Study – CAS(K. A prospective study of knee pain and knee osteoarthritis in the general population

    Directory of Open Access Journals (Sweden)

    Hay Elaine

    2004-02-01

    Full Text Available Abstract Background Knee pain affects an estimated 25% of the adult population aged 50 years and over. Osteoarthritis is the most common diagnosis made in older adults consulting with knee pain in primary care. However, the relationship between this diagnosis and both the current disease-based definition of osteoarthritis and the regional pain syndrome of knee pain and disability is unclear. Expert consensus, based on current evidence, views the disease and the syndrome as distinct entities but the clinical usefulness of these two approaches to classifying knee pain in older adults has not been established. We plan to conduct a prospective, population-based, observational cohort study to investigate the relative merits of disease-based and regional pain syndrome-based approaches to classification and prognosis of knee pain in older adults. Methods All patients aged 50 years and over registered with three general practices in North Staffordshire will be invited to take part in a two-stage postal survey. Respondents to this survey phase who indicate that they have experienced knee pain within the previous 12 months will be invited to attend a research clinic for a detailed assessment. This will consist of clinical interview, physical examination, digital photography, plain x-rays, anthropometric measurement and a brief self-complete questionnaire. All consenting clinic attenders will be followed up by (i general practice medical record review, (ii repeat postal questionnaire at 18-months.

  7. 《阿芒提拉多酒桶》的反讽艺术%Irony in Edgar Allan Poe's The Cask of Amontillado

    Institute of Scientific and Technical Information of China (English)

    李璇; 李晶晶

    2009-01-01

    爱伦·坡具有哥特色彩的著名恐怖短篇小说有效地运用了反讽的叙述手法.通过文本分析,笔者在这篇小说中发现了言语反讽、情景反讽、结构反讽三种反讽类型.这三种反讽的运用产生了显著的美学效果,有效地揭示了爱伦·坡惯于描写的复仇、恐怖、死亡等哥特主题.

  8. Heat Removal Performance in accordance with the Location of the Half-blockage of the Inlet Openings of Concrete Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Yu, S. H.; Lee, J. C.; Lee, S. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The heat transfer rate to the ambient atmosphere by convective air through a passive heat removal system under off-normal conditions reached 87.4 %. Therefore, the half-blockage of the inlet openings has a relatively small effect on the maximum temperature and temperature distributions. A temperature difference in accordance with the location of the half-blockage of the inlet openings was not found. Therefore, the influence of the direction of the half-blockage of the inlet openings reaching the heat removal performance was estimated to be minimal.

  9. A new mode of SAM domain mediated oligomerization observed in the CASKIN2 neuronal scaffolding protein

    KAUST Repository

    Smirnova, Ekaterina

    2016-08-22

    Background: CASKIN2 is a homolog of CASKIN1, a scaffolding protein that participates in a signaling network with CASK (calcium/calmodulin-dependent serine kinase). Despite a high level of homology between CASKIN2 and CASKIN1, CASKIN2 cannot bind CASK due to the absence of a CASK Interaction Domain and consequently, may have evolved undiscovered structural and functional distinctions.

  10. ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-26

    The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs for off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed

  11. Analysis of Spent Nuclear Fuel Imaging Using Multiple Coulomb Scattering of Cosmic Muons

    CERN Document Server

    Chatzidakis, Stylianos; Tsoukalas, Lefteri H

    2016-01-01

    Cosmic ray muons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering. The strong dependence of scattering on atomic number Z and the recent developments on position sensitive muon detectors indicate that multiple Coulomb scattering could be an excellent candidate for spent nuclear fuel imaging. Muons present significant advantages over existing monitoring and imaging techniques and can play a central role in monitoring nuclear waste and spent nuclear fuel stored in dense well shielded containers. The main purpose of this paper is to investigate the applicability of multiple Coulomb scattering for imaging of spent nuclear fuel dry casks stored within vertical and horizontal commercial storage dry casks. Calculations of muon scattering were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fu...

  12. APPLICATION OF MONTE CARLO METHOD IN THE SHIELDING CALCULATION FOR THE SPENTFUEL TRANSPORT CASK%蒙特卡罗方法在乏燃料运输容器屏蔽中的应用

    Institute of Scientific and Technical Information of China (English)

    薛娜; 赵博

    2007-01-01

    本文主要应用蒙特卡罗方法进行了屏蔽计算的研究.在开展蒙特卡罗方法的实际应用中,对大亚湾第一炉换料乏燃料运输容器进行了屏蔽计算.源项应用ORIGEN-Ⅱ程序的计算结果,应用MCNP程序对乏燃料运输容器(STC)进行屏蔽计算,并将计算结果与实测值进行了比较分析.考虑到今后可能会运输更高燃耗的乏燃料,本文对这类乏燃料的运输也进行了计算.

  13. The Classification of Weld Seam Defects for Quantitative Analysis by means of Ultrasonic Testing

    Science.gov (United States)

    Salchak, Y.; Tverdokhlebova, T.; Sharavina, S.; Lider, A.

    2016-06-01

    The paper describes effective quality assessment of spent nuclear fuel storage cask. The ultrasonic testing method is considered. The classification of possible defects with corresponding dimensions limits is proposed. The database of defects of the spent nuclear fuel storage cask was created in compliance with the nuclear energy industry regulatory documents.

  14. Simulation of the dynamic response of radioactive material shipping package-railcar systems during coupling operations

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1983-10-01

    The basic equations of the computer model CARDS (Cask-Railcar Dynamic Simulator), developed for the US Nuclear Regulatory Commission to simulate the dynamic behavior of radioactive material shipping package - railcar systems, are presented. A companion model, CARRS (Cask Railcar Response Spectrum Generator), that generates system response as frequency response spectra is also presented in terms of its basic equations. 1 reference, 18 figures.

  15. Simulation of the dynamic response of radioactive material shipping package - railcar systems during coupling operations

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1981-12-01

    The basic equations of the computer model CARDS (Cask-Railcar Dynamic Simulator), developed for the U.S. Nuclear Regulatory Commission to simulate the dynamic behavior of radioactive material shipping package - railcar systems, are presented. A companion model, CARRS (Casks Railcar Response Spectrum Generator), that generates system response as frequency response spectra is also presented in terms of its basic equations.

  16. GNS Castor V/21 Headspace Gas Sampling 2014

    Energy Technology Data Exchange (ETDEWEB)

    Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    Prior to performing an internal visual inspection, samples of the headspace gas of the GNS Castor V/21 cask were taken on June 12, 2014. These samples were taken in support of the CREIPI/Japanese nuclear industry effort to validate fuel integrity without visual inspection by measuring the 85Kr content of the cask headspace

  17. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  18. Dosimetry at an interim storage for spent nuclear fuel.

    Science.gov (United States)

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons.

  19. Cyclic neutron activation for non-destructive characterization of radioactive waste; Zyklische Neutronen-Aktivierung zur Zerstoerungsfreien Charakterisierung radioaktiver Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Havenith, Andreas; Kettler, John [RWTH Aachen (Germany). Inst. fuer Nuklearen Brennstoffkreislauf; Mauerhofer, Eric [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung

    2011-07-01

    In Germany about 100.000 casks with low-level radioactive waste are actually stored in federal collection sites or at the nuclear power plants. Due to the incomplete documentation these casks have to be characterized with respect to their composition. In order to avoid the opening of the casks a new non-destructive characterization method was developed by the authors based on the prompt and delayed gamma-neutron-activation analyses using 14 MeV neutrons. the main challenge was to determine the self-shielding of neutrons and photons dependent on the sample composition. Computerized MNCP calculations and experiments were performed. The multi-element analysis is based on the gamma spectroscopy during neutron activation. A new measuring system (MEDINA - multi-element detection based on instrumental neutron activation) was built esp. for the characterization of 200-l casks used in the repository KONRAD.

  20. 75 FR 12315 - Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-03-15

    ... MPC-32 canisters and, to allow linear interpolation for some enrichments consistent with the Holtec... conditions in the annular gap between the MPC and the transfer cask depending on which drying process is...

  1. Analytical Chemistry Laboratory (ACL) procedure compendium. Volume 7, Safety operation procedure for hot cell

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    This volume contains the interim change notice for the safety operation procedure for hot cell. It covers the master-slave manipulators, dry waste removal, cell transfers, hoists, cask handling, liquid waste system, and physical characterization of fluids.

  2. 324 Building spent fuel segments pieces and fragments removal summary report

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, C L

    2003-01-09

    As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

  3. Return of spent fuel from the Portuguese research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramalho, A.J.G.; Marques, J.G.; Cardeira, F.M. [Instituto Tecnologico e Nuclear, PO-2686-953 Sacavem (Portugal)

    2000-07-01

    Thirty-nine spent MTR fuel assemblies from the Portuguese Research Reactor were recently returned to the US. Prior to the shipment all assemblies were inspected for corrosion and sipped for determination of fission product leakage. Limitations on the floor loading of the reactor building and on the capacity of the crane prevented the placement and loading of the Transnucleaire IU04 transport cask inside the containment building. The transport cask was thus placed outside, under permanent surveillance, in a support structure built around it. A small transfer cask was used to carry individually the assemblies from the storage racks to the transport cask. A forklift was used as a shuttle between the pool and the IU04. A detailed description of the procedures is given. (author)

  4. Used fuel extended storage security and safeguards by design roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric Richard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jones, Robert [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Ketusky, Edward [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); England, Jeffrey [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Scherer, Carolynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sprinkle, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Michael. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rauch, Eric [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-05-01

    In the United States, spent nuclear fuel (SNF) is safely and securely stored in spent fuel pools and dry storage casks. The available capacity in spent fuel pools across the nuclear fleet has nearly reached a steady state value. The excess SNF continues to be loaded in dry storage casks. Fuel is expected to remain in dry storage for periods beyond the initial dry cask certification period of 20 years. Recent licensing renewals have approved an additional 40 years. This report identifies the current requirements and evaluation techniques associated with the safeguards and security of SNF dry cask storage. A set of knowledge gaps is identified in the current approaches. Finally, this roadmap identifies known knowledge gaps and provides a research path to deliver the tools and models needed to close the gaps and allow the optimization of the security and safeguards approaches for an interim spent fuel facility over the lifetime of the storage site.

  5. Radiation analysis for a generic centralized interim storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, S.G.; Lopez, P. [TRW Environmental Safety Systems, Inc., Vienna, VA (United States); Eble, R.G. [Duke Engineering and Services, Inc., Charlotte, NC (United States)

    1997-12-31

    This paper documents the radiation analysis performed for the storage area of a generic Centralized Interim Storage Facility (CISF) for commercial spent nuclear fuel (SNF). The purpose of the analysis is to establish the CISF Protected Area and Restricted Area boundaries by modeling a representative SNF storage array, calculating the radiation dose at selected locations outside the storage area, and comparing the results with regulatory radiation dose limits. The particular challenge for this analysis is to adequately model a large (6000 cask) storage array with a reasonable amount of analysis time and effort. Previous analyses of SNF storage systems for Independent Spent Fuel Storage Installations at nuclear plant sites (for example in References 5.1 and 5.2) had only considered small arrays of storage casks. For such analyses, the dose contribution from each storage cask can be modeled individually. Since the large number of casks in the CISF storage array make such an approach unrealistic, a simplified model is required.

  6. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Hostick, C.J. [Pacific Northwest Lab., Richland, WA (United States); Lavender, J.C. [Westinghouse Hanford Co., Richland, WA (United States); Wakeman, B.H. [Virginia Electric and Power Co., Richmond, VA (United States)

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel.

  7. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    Energy Technology Data Exchange (ETDEWEB)

    Sicard, D.; Verdier, A. [COGEMA Logistics (AREVA Group) (France); Monsigny, P.A. [NOK/KKL (Switzerland)

    2004-07-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators.

  8. Data and methods for the assessment of the risks associated with the maritime transport of radioactive materials: Results of the SeaRAM program studies. Volume 2 -- Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; Bespalko, S.J.; Kanipe, F.L. [and others

    1998-05-01

    This report describes ship accident event trees, ship collision and ship fire frequencies, representative ships and shipping practices, a model of ship penetration depths during ship collisions, a ship fire spread model, cask to environment release fractions during ship collisions and fires, and illustrative consequence calculations. This report contains the following appendices: Appendix 1 -- Representative Ships and Shipping Practices; Appendix 2 -- Input Data for Minorsky Calculations; Appendix 3 -- Port Ship Speed Distribution; and Appendix 4 -- Cask-to-Environment Release Fractions.

  9. Overweight truck shipments to nuclear waste repositories: legal, political, administrative and operational considerations

    Energy Technology Data Exchange (ETDEWEB)

    1986-03-01

    This report, prepared for the Chicago Operations Office and the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE), identifies and analyzes legal, political, administrative, and operational issues that could affect an OCRWM decision to develop an overweight truck cask fleet for the commercial nuclear waste repository program. It also provides information required by DOE on vehicle size-and-weight administration and regulation, pertinent to nuclear waste shipments. Current legal-weight truck casks have a payload of one pressurized-water reactor spent fuel element or two boiling-water reactor spent fuel elements (1 PWR/2 BWR). For the requirements of the 1960s and 1970s, casks were designed with massive shielding to accommodate 6-month-old spent fuel; the gross vehicle weight was limited to 73,280 pounds. Spent fuel to be moved in the 1990s will have aged five years or more. Gross vehicle weight limitation for the Interstate highway system has been increased to 80,000 pounds. These changes allow the design of 25-ton legal-weight truck casks with payloads of 2 PWR/5 BWR. These changes may also allow the development of a 40-ton overweight truck cask with a payload of 4 PWR/10 BWR. Such overweight casks will result in significantly fewer highway shipments compared with legal-weight casks, with potential reductions in transport-related repository risks and costs. These advantages must be weighed against a number of institutional issues surrounding such overweight shipments before a substantial commitment is made to develop an overweight truck cask fleet. This report discusses these issues in detail and provides recommended actions to DOE.

  10. Localization and monitoring of spent fuel containers applying electromagnetic reflection measurement (EMR). Final report; Ortung und Ueberwachung von Brennelementbehaeltern mit elektromagnetischen Reflexionsmessungen (EMR). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The direct final disposal of spent nuclear fuels involves the emplacement in containers, e.g. Pollux casks, and their permanent disposal in drifts. The IAEA requires surveillance measures for this concept. By the BGR the electromagnetic reflection method (EMR, underground radar) has been suggested for surveillance. It was tested for its suitability in the Asse salt mine on a rock-up of Pollux casks. (DG)

  11. A robotized surface workstation for manipulation, filling and closing of packaging containers for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bartos, Pavel [FITE a.s., Ostrava-Marianske Hory (Czech Republic); Haladova, Petra [Robotsystem, LLC/Moravian Research, LLC, Ostrava-Moravska (Czech Republic); Otcenasek, Petr

    2016-01-15

    Options for the handling of spent nuclear fuel are described and a packaging cask for an underground repository is presented as also a robotic surface workplace for the repository. The potential for the closing the nuclear fuel cycle is discussed. Currently, a team of Czech experts is developing a project of fully robotic technology for manipulation and storage of packaging casks for spent nuclear fuel in host rock of underground repository.

  12. RH Packaging Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2003-09-17

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

  13. Aluminum honeycomb impact limiter study

    Energy Technology Data Exchange (ETDEWEB)

    Yaksh, M.C.; Thompson, T.C. (Nuclear Assurance Corp., Norcross, GA (United States)); Nickell, R.E. (Applied Science and Technology, Inc., Poway, CA (United States))

    1991-07-01

    Design requirements for a cask transporting radioactive materials must include the condition of the 30-foot free fall of the cask onto an unyielding surface. To reduce the deceleration loads to a tolerable level for all the components of the cask, a component (impact limiter) is designed to absorb the kinetic energy. The material, shape, and method of attachment of the impact limiter to the cask body comprises the design of the impact limiter. The impact limiter material of interest is honeycomb aluminum, and the particular design examined was for the NAC Legal Weight Truck cask (NAC-LWT) for spent fuel from light water reactors. The NAC-LWT has a design weight of 52,000 pounds, and it has a nominal length of 200 inches. The report describes the numerical calculations embodied in the FADE program to determine the accelerations and crush strain resulting from an arbitrary height and angle of orientation. Since the program serves as a design tool, static tests are performed to assess the effect of the shell containing the honeycomb aluminum. The static tests and their results are contained in the study. The static tests are used to demonstrate for licensing purposes the level of accelerations imposed on the cask during a 30-foot drop. 3 refs., 41 figs., 15 tabs.

  14. Study of the effect of gamma irradiation ({sup 60} Co) on the quality of sugar cane spirit and on the cask of maturation; Estudo do efeito da irradiacao gamma ({sup 60} Co) na qualidade da cachaca e no tonel de envelhecimento

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, Mariana Branco de; Horii, Jorge; Alcarde, Andre Ricardo [Universidade de Sao Paulo (USP), Piracicaba, SP (Brazil). Escola Superior de Agricultura Luiz de Queiroz (ESALQ). Dept.de Agroindustria, Alimentos e Nutricao]. E-mails: jhorii@esalq.usp.br; aralcard@esalq.usp.br

    2006-10-15

    The ageing or maturing process of drink improves the sensorial characteristics of the product making it a qualified drink with higher economical value. The traditional maturing method of drinks involves their interaction with wood and irradiation can accelerate this ageing process. Sugar cane spirit and oak barrels with a 20 L capacity were treated using gamma irradiation (150 Gy). Physical-chemical and chromatographic analyses were constantly performed for 390 days during the drink ageing period. The sugar cane spirit and barrel irradiation did not change most volatile components of the same type coefficient such as volatile acidity, esters, superior alcohols and furfural throughout the 390 days. There is evidence, however, that some component parameters like aldehydes, tannin, color and copper concentration are in some way influenced resulting in partial acceleration of the ageing or maturing process. At the end of the ageing period, a sensorial analysis was made using 30 non-trained people who tasted the drink. The ageing process acceleration was confirmed by the sensorial evaluation, where the irradiated sugar cane spirit and/or barrel had higher approval regarding all parameters analyzed (scent, flavor and appearance). (author)

  15. Estudo do efeito da irradiação gamma (60CO na qualidade da cachaça e no tonel de envelhecimento Study of the effect of gamma irradiation (60co on the quality of sugar cane spirit and on the cask of maturation

    Directory of Open Access Journals (Sweden)

    Mariana Branco de Miranda

    2006-12-01

    Full Text Available O processo de envelhecimento ou maturação das bebidas proporciona uma melhora nas características sensoriais da cachaça, tornando-a de qualidade superior e de maior valor econômico. O método tradicional de maturação de bebidas é sua interação com madeiras, sendo que a irradiação pode acelerar este processo de envelhecimento. A cachaça e os tonéis de carvalho de 20 L de capacidade foram submetidos à irradiação gama (150 Gy. Análises físico-químicas e cromatográficas foram realizadas periodicamente ao longo de 390 dias do período de envelhecimento da bebida. A irradiação da cachaça e do tonel não alterou a maioria dos componentes voláteis do coeficiente de congêneres como acidez volátil, ésteres, álcoois superiores e furfural durante os 390 dias. Há evidências, entretanto, de que os parâmetros de alguns componentes como aldeídos, taninos, cor e teor de cobre são de alguma forma influenciados, resultando em aceleração parcial do processo de maturação ou envelhecimento. Ao final do período de envelhecimento, foi feita uma análise sensorial com 30 provadores não treinados. A aceleração do processo de envelhecimento foi confirmada pela avaliação sensorial, e a cachaça e/ou tonel irradiados receberam maior indicação de aprovação em todos os parâmetros analisados (aroma, sabor e aparência.The ageing or maturing process of drink improves the sensorial characteristics of the product making it a qualified drink with higher economical value. The traditional maturing method of drinks involves their interaction with wood and irradiation can accelerate this ageing process. Sugar cane spirit and oak barrels with a 20 L capacity were treated using gamma irradiation (150 Gy. Physical-chemical and chromatographic analyses were constantly performed for 390 days during the drink ageing period. The sugar cane spirit and barrel irradiation did not change most volatile components of the same type coefficient such as volatile acidity, esters, superior alcohols and furfural throughout the 390 days. There is evidence, however, that some component parameters like aldehydes, tannin, color and copper concentration are in some way influenced resulting in partial acceleration of the ageing or maturing process. At the end of the ageing period, a sensorial analysis was made using 30 non-trained people who tasted the drink. The ageing process acceleration was confirmed by the sensorial evaluation, where the irradiated sugar cane spirit and/or barrel had higher approval regarding all parameters analyzed (scent, flavor and appearance.

  16. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    CARRELL, R D

    2002-07-16

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

  17. Dry spent fuel storage with the MACSTOR system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations

    1996-10-01

    Atomic Energy of Canada Limited (AECL), and Transnuclear Inc. (TNI) began in 1989 the development of the concrete spent fuel storage system, called MACSTOR (Modular Air-Cooled Canister STORage) for use with LWR spent fuel assemblies. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The MACSTOR Module is a monolithic, shielded concrete vault structure that can accommodate up to 20 spent fuel canisters. Each canister typically holds up to 21 PWR or 44 BWR spent fuel assemblies with a nominal fuel burn up rate of 40,000 MWD/MTU and a 7 year minimum cooling period. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. Thus, the utility can be assured of both positive cooling of the fuel and verification of the integrity of the fuel confinement boundary. The structure is seismically designed and is capable of withstanding site design basis accident events. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. The MACSTOR system can economically address a wide range of storage capacity requirements. The modular concept allows for flexibility in determining each module`s capacity. Starting with 8 canisters, the capacity can be increased by increments of 4 up to 20 canisters. The MACSTOR system is also flexible in accommodating the various spent fuel types from such reactors as VVER-440, VVER-1000 and RBMK 1500. (J.P.N.)

  18. Development of Remote Operations Technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sub; Kwon, Sung Gyu; Kim, Ki Ho; Park, Byung Suk; Park Young Soo; Lee, Won Sang [Korea Atomic Energry Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    To control the swing of the load suspended to overhead crane, various devices, anti-swing control algorithms, and experimental and pilot-scaled cranes have been developed since 1989. In this year, the problems of control algorithm are identified and new algorithms which adopt an acceleration profile planning, a velocity feedback control, a fuzzy control, and an anti-swing/position control methods are developed. Also, the report includes the program of transferring the anti-swing overhead crane technology to industry. And, to improve the cask handling safety and to increase the cask handling capacity of a facility, a Remote Cask Grappling and Lid Unbolting Device (RCG=LUD) which is suspended to an anti-swing crane and remotely controlled to hold the cask and unbolt the cask lid is developed. the controller using PMAC is tested for successful remote operations. Technology for this device has the potential for developing other technologies, such as nuclear reactor maintenance and repair. 62 refs., 73 figs., 27 tabs.

  19. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  20. Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

    Science.gov (United States)

    Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

    Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

  1. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time.

  2. Current developments in container design testing for the Konrad repository

    Energy Technology Data Exchange (ETDEWEB)

    Voelzke, Holger; Nieslony, Gregor; Noack, Volker; Hagenow, Peter; Kovacs, Oliver [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2012-05-15

    In 2002, the Konrad repository was licensed as a repository for radioactive waste generating no heat. That permit subsequently became the object of litigation and was confirmed by a court of last resort as late as in 2007. The Federal Office of Radiation Protection (BfS) then started planning and converting the former iron ore mine into a repository. The licensed repository volume is 303,000 m{sup 3} based on estimates of expected waste arisings. The mine proper would offer a much larger volume. However, cask emplacement can be started only after completion of the repository which, according to the present status, will not be before the end of this decade. Nevertheless, there is great interest even now in conditioning and packaging for repository storage of the radioactive waste planned for Konrad, which also requires casks type tested by the Federal Institute of Materials Testing (BAM) and approved by the BfS. The key items in the license for the repository are comprehensive requirements to be met by waste forms and casks. As far as continuous revision of repository requirements and consideration of materials hazardous to water are concerned, it is assumed that the key requirements applying to type tests of casks with respect to waste forms and casks will be affected by this either not at all or only very slightly. (orig.)

  3. Expression of the human TSPY gene in the brains of transgenic mice suggests a potential role of this Y chromosome gene in neural functions

    Institute of Scientific and Technical Information of China (English)

    Tatsuo Kido; Stephanie Schubert; J(o)rg Schmidtke; Yun-Fai Chris Lau

    2011-01-01

    The testis specific protein Y-encoded (TSPY) is a member of TSPY/SET/NAPl superfamily, encoded within the gonadoblastoma locus on the Y chromosome. TSPY shares a highly conserved SET/NAP-domain responsible for protein-protein interaction among TSPY/SET/NAPl proteins.Accumulating data, so far, support the role of TSPY as the gonadoblastoma gene, involved in germ cell tumorigenesis. The X-chromosome homolog of TSPY, TSPX is expressed in various tissues at both fetal and adult stages, including the brain, and is capable of interacting with the multi-domain adapter protein CASK, thereby influencing the synaptic and transcriptional functions and developmental regulation of CASK in the brain and other neural tissues. Similar to TSPX, we demonstrated that TSPY could interact with CASK at its SET/NAP-domain in cultured cells. Transgenic mice harboring a human TSPY gene and flanking sequences showed specific expression of the human TSPYtransgene in both testis and brain. The neural expression pattern of the human TSPY gene overlapped with those of the endogenous mouse Cask and Tspx gene. Similarly with TSPX, TSPY was co-localized with CASK in neuronal axon fibers in the brain, suggesting a potential role(s) of TSPY in development and/or physiology of the nervous system.

  4. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  5. Calcium channel structural determinants of synaptic transmission between identified invertebrate neurons.

    Science.gov (United States)

    Spafford, J David; Munno, David W; Van Nierop, Pim; Feng, Zhong-Ping; Jarvis, Scott E; Gallin, Warren J; Smit, August B; Zamponi, Gerald W; Syed, Naweed I

    2003-02-01

    We report here that unlike what was suggested for many vertebrate neurons, synaptic transmission in Lymnaea stagnalis occurs independent of a physical interaction between presynaptic calcium channels and a functional complement of SNARE proteins. Instead, synaptic transmission in Lymnaea requires the expression of a C-terminal splice variant of the Lymnaea homolog to mammalian N- and P/Q-type calcium channels. We show that the alternately spliced region physically interacts with the scaffolding proteins Mint1 and CASK, and that synaptic transmission is abolished following RNA interference knockdown of CASK or after the injection of peptide sequences designed to disrupt the calcium channel-Mint1 interactions. Our data suggest that Mint1 and CASK may serve to localize the non-L-type channels at the active zone and that synaptic transmission in invertebrate neurons utilizes a mechanism for optimizing calcium entry, which occurs independently of a physical association between calcium channels and SNARE proteins.

  6. ENSA, loading containers: operational experience as the basis for a vital future; Ensa, cargando contenedores: la experiencia operativa como base de un futuro vital

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, H.; Sotto, A.

    2015-07-01

    Used Fuel Management have turned itself a vital necessity for the nuclear power plants for a short and mid-term, especially for those in its halfway. Ensa, with a proved experience over 50 cask loading, is now ready to make the step forward and load its first new design Ensa Universal Cask (ENUN) in the incoming year. After the experience carried out with the reference cask DPT-21, is also time to jump into the continuous improvement and where all the lessons learned can be applied in benefit of a new generation of auxiliary tools. Through this post, we try to explain our experience keys and how the technology progress contributes towards the effectiveness with the personnel safety point of view as a reference. (Author)

  7. Literature Survey of Copper Corrosion Modelling under Deep Geological Disposal Environment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo; Kook, Dong-Hak; Lee, Min-Soo; Choi, Heui-Joo

    2008-03-15

    As a one of the solution for the spent nuclear fuel problem, deep repository research have been under way in KAERI. To raise the resistance against repository cask corrosion, copper canister was adopted for the outer material of the cask. Duration of the repository with million year order makes the copper corrosion research under the repository environment very important and the corrosion modelling which could trace the real world precisely and predict the corrosion behavior very well is subsequently necessary. With in progress of plan to manufacture unique copper cask corresponding to our country repository system, survey for the preceding research papers in recent research direction and kernel points is expected to significant. This paper arranged the representative literatures for the important corrosion mechanism which are recently published.

  8. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  9. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  11. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

    Energy Technology Data Exchange (ETDEWEB)

    HOLLENBECK, R.G.

    2000-05-08

    The Spent Nuclear Fuel (SNF) Canister Storage Building (CSB) is the interim storage facility for the K-Basin SNF at the US. Department of Energy (DOE) Hanford Site. The SNF is packaged in multi-canister overpacks (MCOs). The MCOs are placed inside transport casks, then delivered to the service station inside the CSB. At the service station, the MCO handling machine (MHM) moves the MCO from the cask to a storage tube or one of two sample/weld stations. There are 220 standard storage tubes and six overpack storage tubes in a below grade reinforced concrete vault. Each storage tube can hold two MCOs.

  12. Radiation doses from radiation sources of neutrons and photons by different computer calculation; Tecniche di calcolo di intensita` di dose da sorgenti di radiazione neutronica e fotonica con l`uso di codici basati su metodologie diverse

    Energy Technology Data Exchange (ETDEWEB)

    Siciliano, F.; Lippolis, G.; Bruno, S.G. [ENEA, Centro Ricerche Trisaia, Rotondella (Italy)

    1995-11-01

    In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoretic background are concerned.

  13. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); LeDuc, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-25

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

  14. The radiation characteristics of the transport packages with vitrified high-level waste

    Science.gov (United States)

    Bogatov, S. A.; Mitenkova, E. F.; Novikov, N. V.

    2015-12-01

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  15. The radiation characteristics of the transport packages with vitrified high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Bogatov, S. A. [JSC VNIPIpromtechnologii (Russian Federation); Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

    2015-12-15

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  16. Signatures of Extended Storage of Used Nuclear Fuel Comprehensive Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-21

    This report serves as a comprehensive overview of the Extended Storage of Used Nuclear Fuel work performed for the Material Protection, Accounting and Control Technologies campaign under the Department of Energy Office of Nuclear Energy. This paper describes a signature based on the source and fissile material distribution found within a population of used fuel assemblies combined with the neutron absorbers found within cask design that is unique to a specific cask with its specific arrangement of fuel. The paper describes all the steps used in producing and analyzing this signature from the beginning to the project end.

  17. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-24

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

  18. Construction and Perfection of the Management Mechanism of Competitive Sports from the Perspective of Cask Effect Theory——Thinking of the Phenomenon of Chinese Short Track Speed Skating Team%“木桶理论”视角下的中国竞技体育管理机制的构建与完善——对中国短道速滑队现象的思考

    Institute of Scientific and Technical Information of China (English)

    李丽

    2012-01-01

    中国短道速滑队事件暴露出中国竞技体育管理机制上的漏洞与缺失.为实现中国由体育大国向着体育强国嬗变的战略目标,建立健全竞技体育管理机制,实现竞技体育的科学化管理,是当前我国体育界必须给予高度重视并切实加以解决的重要课题.运用“木桶理论”对我国竞技体育管理机制的构建进行论述,旨在为实现我国竞技体育健康持久的发展,提供必要的理论支撑与参考.

  19. SCALE5.1程序系统中蒙特卡罗方法模块与离散纵坐标模块在乏燃料运输容器屏蔽计算中的比较分析%A Comparison of Shielding Calculation for a Spent Fuel Shipping Cask by Using 3D Monte Carlo and 1D Discrete-ordinates Methods

    Institute of Scientific and Technical Information of China (English)

    张普忠; 陈义学; 毛亚蔚; 马续波; 韩静茹; 杨寿海; 石生春; 张斌; 王继亮

    2010-01-01

    以某型乏燃料运输容器为计算模型,分别利用SCALE5.1程序系统中的一维离散纵标法程序和三维蒙特卡罗方法程序对运输容器进行了屏蔽计算,计算结果表明,两种方法的总当量剂量率结果相对偏差在10%以内.最后对两个模块的应用特点及差异进行了比较分析,为其在乏燃料容器屏蔽计算中的应用提供参考.

  20. 10 CFR 72.48 - Changes, tests, and experiments.

    Science.gov (United States)

    2010-01-01

    ... facility or spent fuel storage cask design, of changes in procedures, and of tests and experiments made... 10 Energy 2 2010-01-01 2010-01-01 false Changes, tests, and experiments. 72.48 Section 72.48... Issuance and Conditions of License § 72.48 Changes, tests, and experiments. (a) Definitions for...

  1. 77 FR 12887 - Virgina Electric and Power Company; Receipt of Request for Action

    Science.gov (United States)

    2012-03-02

    ... to keep North Anna 1 and 2, in a ``cold shutdown'' mode of operation until such time as a series of... Anna 1 and 2, Independent Spent Fuel Storage Installation (ISFSI) including the fact that 25 casks... is available for inspection at the Commission's Public Document Room, located at One White...

  2. 10 CFR 75.4 - Definitions.

    Science.gov (United States)

    2010-01-01

    ... graphite, irradiated fuel casks and canisters, reactor control rods, criticality safe tanks and vessels...; reactors; critical facilities; reprocessing of nuclear fuel; and processing of intermediate or high-level... nuclear material, such as an independent spent fuel storage installation (ISFSI) or a...

  3. PATRAM '80. Proceedings. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Huebner, H.W. (ed.)

    1980-01-01

    Volume 2 contains papers from the following sessions: Safeguards-Related Problems; Neutronics and Criticality; Operations and Systems Experience II; Plutonium Systems; Intermediate Storage in Casks; Operations and Systems Planning; Institutional Issues; Structural and Thermal Evaluation I; Poster Session B; Extended Testing I; Structural and Thermal Evaluation II; Extended Testing II; and Emergency Preparedness and Response. Individual papers were processed. (LM)

  4. Constructing Complexity: Using Reading Levels to Differentiate Reading Comprehension Activities

    Science.gov (United States)

    FitzPatrick, Declan

    2008-01-01

    The author remembers a class when he asked his students to discuss in small groups how Edgar Allan Poe suggests a judgment of the main character in "The Cask of Amontillado". During their discussion it became clear to the author that the students couldn't come to consensus because they had no grasp of the narrator's explanations of his motivations…

  5. 75 FR 36449 - Yankee Atomic Electric Co.; Yankee Atomic Independent Spent Fuel Storage Installation; Issuance...

    Science.gov (United States)

    2010-06-25

    ... Transportation, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington...-MPC, Certificate of Compliance (CoC) No. 1025, to store spent nuclear fuel under a general license in... Nuclear Power Station, located at Rowe, Massachusetts. YAEC stores spent fuel in fifteen NAC-MPC casks...

  6. Latin America Report, No. 2675.

    Science.gov (United States)

    2007-11-02

    that the benefits to be derived from membership of the grouping, headed by India’s Prime Minister Indira Gandhi , were "primarily diplomatic." He also...inhabitants of Ciudad Lopez Mateos. Our informants expressed their ill feeling towards the owners of the "casks" some of whom "load their cistern carts in

  7. Oak Ridge National Laboratory shipping containers for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Schaich, R.W.

    1980-05-01

    The types of containers used at ORNL for the transport of radioactive materials are described. Both returnable and non-returnable types are included. Containers for solids, liquids and gases are discussed. Casks for the shipment of uranium, irradiated fuel elements, and non-irradiated fuel elements are also described. Specifications are provided. (DC)

  8. Experience of IEA-R1 research reactor spent fuel transportation back to United States

    Energy Technology Data Exchange (ETDEWEB)

    Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Operacao do Reator IEAR-R1m]. E-mail: frajndli@net.ipen.br; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div.de Engenharia do Nucleo]. E-mail: perrotta@net.ipen.br; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Diretoria de Reatores]. E-mail: maiorino@net.ipen.br; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores]. E-mail: ajsoares@net.ipen.br

    1998-07-01

    IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)

  9. United States Army Advanced Medic (91B30) Training: An Iterative Decision Method Application

    Science.gov (United States)

    1983-03-25

    be provided through subsequent boards. Croup casks, group Interaccion process and group performance effecciveness: A review and In...L. Berkowlcz (Ed.), Advances In Experimental Social Psychology (Vol 8). New York: Ac- Reference Noce I. Flnstuen, K...discussion upon the correctness of group decisions, when the factor of majority influ- ence is allowed for. Journal of Social

  10. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains & Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer.

  11. Ultrasonic fingerprinting by phased array transducer

    Science.gov (United States)

    Sednev, D.; Kataeva, O.; Abramets, V.; Pushenko, P.; Tverdokhlebova, T.

    2016-06-01

    Increasing quantity of spent nuclear fuel that must be under national and international control requires a novel approach to safeguard techniques and equipment. One of the proposed approaches is utilize intrinsic features of casks with spent fuel. In this article an application of a phased array ultrasonic method is considered. This study describes an experimental results on ultrasonic fingerprinting of austenitic steel seam weld.

  12. Ultrasonic fingerprinting by phased array transducer

    OpenAIRE

    Sednev, Dmitry Andreevich; Kataeva, O.; Abramets, V.; Pushenko, P.; Tverdokhlebova, T.

    2016-01-01

    Increasing quantity of spent nuclear fuel that must be under national and international control requires a novel approach to safeguard techniques and equipment. One of the proposed approaches is utilize intrinsic features of casks with spent fuel. In this article an application of a phased array ultrasonic method is considered. This study describes an experimental results on ultrasonic fingerprinting of austenitic steel seam weld.

  13. Instrumentation: Nondestructive Examination for Verification of Canister and Cladding Integrity. FY2014 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-12

    This report documents FY14 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) verify the integrity of dry storage cask internals.

  14. Evaluation of Simplified Models for Estimating Public Dose from Spent Nuclear Fuel Shipments

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Kevin J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This paper investigates the dose rate as a function of distance from a representative high-capacity SNF rail-type transportation cask. It uses the SCALE suite of radiation transport modeling and simulation codes to determine neutron and gamma radiation dose rates. The SCALE calculated dose rate is compared with simplified analytical methods historically used for these calculations. The SCALE dose rate calculation presented in this paper employs a very detailed transportation cask model (e.g., pin-by-pin modeling of fuel assembly) and a new hybrid computational transport method. Because it includes pin-level heterogeneity and models ample air and soil outside the cask to simulate scattering of gamma and neutron radiation, this detailed SCALE model is expected to yield more accurate results than previously used models which made more simplistic assumptions (e.g., fuel assembly treated as a point or line source, simple 1-D model of environment outside of cask). The results in this paper are preliminary and, as progress is made on developing and validating improved models, results may be subject to change as models and estimates become more refined and better information leads to more accurate assumptions.

  15. Evaluation of Transportation Options for Intermediate Non-destructive Examinations

    Energy Technology Data Exchange (ETDEWEB)

    Case, Susan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoggard, Gary [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-11-01

    Idaho National Laboratory (INL) shipments of irradiated experiments from the Advanced Test Reactor (ATR) to the Hot Fuels Examination Facility (HFEF) have historically been accomplished using the General Electric Model 2000 (GE 2000) Type B shipping container. Battelle Energy Alliance (BEA) concerns regarding the future availability and leasing and handling costs associated with the GE 2000 cask have warranted an evaluation of alternative shipping options. One or more of these shipping options may be utilized to perform non-destructive examinations (NDE) such as neutron radiography and precision gamma scans of irradiated experiments at HFEF and then return the experiments to ATR for further irradiation, hereafter referred to as “intermediate NDE.” This evaluation includes transportation options for intermediate NDE using the GE 2000 cask, BEA Research Reactor (BRR) package, Dry Transfer Cubicle (DTC) insert, and the General Electric Model 100 (GE 100) cask. The GE 2000 cask is the only Type B shipping container currently in use for shipments of irradiated material (exceeding Type A quantities) from ATR to HFEF; therefore it is included as one of the four shipping options in this evaluation. Cost and schedule estimates are provided for performing neutron radiography and precision gamma scans of a five-capsule drop-in-type ATR experiment for each transportation option. All costs provided in this evaluation are rough order-of-magnitude costs based on input from knowledgeable vendor employees and individuals at INL facilities.

  16. Data and methods for the assessment of the risks associated with the maritime transport of radioactive materials: Results of the SeaRAM program studies. Volume 1 -- Main report

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; Bespalko, S.J.; Kanipe, F.L. [and others

    1998-05-01

    This report describes ship accident event trees, ship collision and ship fire frequencies, representative ships and shipping practices, a model of ship penetration depths during ship collisions, a ship fire spread model, cask to environment release fractions during ship collisions and fires, and illustrative consequence calculations.

  17. 75 FR 18242 - Florida Power and Light, St. Lucie Units 1 and 2; Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-04-09

    ... the human environment. Accordingly, the Commission has determined that an ] environmental impact... assemblies, complicating the fuel handling evolutions required for core reload during the Unit 1 RFO. In... casks and does not affect non-radiological plant effluents, or any other aspects of the...

  18. Preliminary risk assessment for nuclear waste disposal in space, volume 2

    Science.gov (United States)

    Rice, E. E.; Denning, R. S.; Friedlander, A. L.

    1982-01-01

    Safety guidelines are presented. Waste form, waste processing and payload fabrication facilities, shipping casks and ground transport vehicles, payload primary container/core, radiation shield, reentry systems, launch site facilities, uprooted space shuttle launch vehicle, Earth packing orbits, orbit transfer systems, and space destination are discussed. Disposed concepts and risks are then discussed.

  19. 27 CFR 31.231 - Destruction of marks and brands on wine containers.

    Science.gov (United States)

    2010-04-01

    ... brands on wine containers. 31.231 Section 31.231 Alcohol, Tobacco Products and Firearms ALCOHOL AND... § 31.231 Destruction of marks and brands on wine containers. A dealer who empties any cask, barrel, keg, or other bulk container of wine must scrape or obliterate from the empty container all marks,...

  20. Neutron spectrometry at the interim storage facility for spent nuclear fuel

    CERN Document Server

    Králik, M; Studeny, J

    2002-01-01

    Dosimetric characteristics of neutron and photon components of mixed fields around casks for spent nuclear fuel have been determined at various places at the dry interim storage facility. The results obtained with metrological grade instruments were compared with data provided by usual survey meters for both neutrons and photons.

  1. 27 CFR 25.35 - Tanks.

    Science.gov (United States)

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Tanks. 25.35 Section 25.35... TREASURY LIQUORS BEER Construction and Equipment Equipment § 25.35 Tanks. Each stationary tank, vat, cask... contents of tanks or containers in lieu of providing each tank or container with a measuring device....

  2. Simulation of shock absorbing materials loaded by drop weights

    Energy Technology Data Exchange (ETDEWEB)

    Kasparek, Eva; Voelzke, Holger; Scheidemann, Robert [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2013-08-15

    The numerical simulation of transport and storage casks for radioactive waste in accidental scenarios requires adequate material models and input parameters for all relevant structural cask elements, the impact foundation as well as possibly further components involved in the incident. They form the basis for reliable analysis of different loading situations allowing substituting, at least in parts, experimental investigations that are needed in safety reports submitted for nuclear licensing procedures. Hereby, the efficiency of the energy absorbing components, such as cask impact limiters or equivalent structures placed onto or into the foundation of storage buildings, determine to a large extend the performance and thus the design of the casks. The limiters, which are usually made of wood, polyurethane foam (PUF) or damping concrete, are subjected to major plastic deformations, whose distribution and magnitude strongly depend on loading speed, course and temperature. Consequently, it is necessary to establish a systematic data basis and implement this information in advanced numerical material formulations in order to enable finite element simulations to account for all relevant factors. Since the end of 2008, BAM (Bundesanstalt fuer Materialforschung und -pruefung) is supported by the Federal Ministry of Education and Research in order to determine such experimental data as well as to adapt and develop qualified material models for shock absorbing materials in the framework of a project called ENREA. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-18

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

  4. Remote-Handled Transuranic Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2001-08-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document representsthe development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specificidentification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing|description for each content code and compiles this information for all DOE sites. Compliance withwaste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The contentcode gives an overall description of the RH-TRU waste material in terms of processes and|packaging, as well as the generation location. This helps to provide cradle-to-grave traceability ofthe waste material so that the various actions required to assess its qualification as payload for the72-B cask can be performed. The content codes also impose restrictions and requirements on themanner in which a payload can be assembled.The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures|applicable for the qualification of waste as payload for the 72-B cask. The logic for this|classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON,|RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and|justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. |Each content code uniquely

  5. CASTOR GSF packaging design criteria

    Energy Technology Data Exchange (ETDEWEB)

    Burnside, M.E.

    1996-08-06

    Encapsulated vitrified materials (Isotopic Heat Sources) are currently stored in the Pacific Northwest National Laboratories (PNNL) 324 Building located in the 300 Area. As part of the 324 Building transition program, the vitrified material, encapsulated in stainless steel canisters, must be removed. These canisters were originally intended to be used by the German government, but are no longer desired. As part of an agreement with the German government, the Germans are providing the U.S. Department of Energy (DOE) with six (6) CASTOR GSF and four (4) GNS-12 casks.The canisters will be transported onsite in CASTOR GSF and GNS-12 casks for interim storage until final disposition of the material is determined.

  6. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  7. MICrocephaly, disproportionate pontine and cerebellar hypoplasia syndrome: A clinico-radiologic phenotype linked to calcium/calmodulin-dependent serine protein kinase gene mutation

    Directory of Open Access Journals (Sweden)

    Rashid Saleem

    2013-01-01

    Full Text Available MICrocephaly, disproportionate pontine and cerebellar hypoplasia (MICPCH syndrome, a rare X-linked disorder, generally seen in girls, is characterized by neurodevelopmental delay, microcephaly, and disproportionate pontine and cerebellar hypoplasia. It is caused by inactivating calcium/calmodulin-dependent serine protein kinase (CASK gene mutations. We report a 2-year-old girl with severe neurodevelopmental delay, microcephaly, minimal pontine hypoplasia, cerebellar hypoplasia, and normal looking corpus callosum, with whom the conventional cytogenetic studies turned out to be normal, and an array-comparative genomic hybridization (a-CGH analysis showed CASK gene duplication at Xp11.4. Our case highlights the importance of using clinico-radiologic phenotype to guide genetic investigation and it also confirms the role of a-CGH analysis in establishing the genetic diagnosis of MICPCH syndrome, when conventional cytogenetic studies are inconclusive.

  8. AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Dewes, J.

    2014-02-24

    The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

  9. Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

  10. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  11. Rigid performance requirements assure public safety by regulations

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E.; Glass, R.E.

    1987-07-01

    Title 10, Part 71 of the Code of Federal Regulations (10 CFR 71) provides a set of prescriptive performance test requirements for spent nuclear fuel and high-level waste transport packaging containment systems. The hypothetical accident conditions, which involve a sequence of impact, puncture, fire, and water immersion events, are referred to as rigid because of their extremely prescriptive nature. These hypothetical accident events have now been placed within the context of real transportation accidents, at least for conventional austenitic stainless steel/ lead gamma shielded cask designs. The assurance of public safety, including the issue of safety margin for very severe accident events is discussed in this paper for both conventional and innovative cask design concepts. A particular risk assessment approach that follows from work at the Lawrence Livermore National Laboratory is suggested.

  12. Rigid performance requirements assure public safety by regulations

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E.; Glass, R.E.

    1987-01-01

    Title 10, Part 71 of the Code of Federal Regulations (10 CFR 71) provides a set of prescriptive performance test requirements for spent nuclear fuel and high-level waste transport packaging containment systems. The hypothetical accident conditions, which involve a sequence of impact, puncture, fire, and water immersion events, are referred to as rigid because of their extremely prescriptive nature. These hypothetical accident events have now been placed within the context of real transportation accidents, at least for conventional austenitic stainless steel/lead gamma shielded cask designs. The assurance of public safety, including the issue of safety margin for very severe accident events, is discussed in this paper for both conventional and innovative cask design concepts. A particular risk assessment approach that follows from work at the Lawrence Livermore National Laboratory is suggested. 6 refs., 2 figs.

  13. Statement of work for the immobilized high-level waste transportation system, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Mouette, P.

    1998-06-24

    The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

  14. Tutorial Workshop on Robotics and Robot Control.

    Science.gov (United States)

    1982-10-26

    would consider the interactive manipulator control as a cask planning and decomposition problem. (See, e.g., Refs. 1 - 6.) That concept usually...interaction with the control system as an essentially off-line and high level programming action. The task planning / decomposition concepc is...distance of the end effeccor from a given surface or object, centering on an object, ect. Complex accions consist of composite actions that are executed

  15. Operation and maintenance techniques of cranes and Hoist in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Woong Sup; Oh, Yon Woo; Ahn, Sang Bock; Park, Dae Gue; Whang, Ryoung Wha

    1999-07-01

    Crane and hoist of 11 sets are installed int the Irradiated Material Examination Facility (IMEF). IMEF is divided into two parts such as hot cell area and service area. 30/5-ton overhead crane is installed in service area for transfer of irradiated material transportation cask and other several kinds of heavy equipment. This report describes maintenance techniques, repair procedure, daily and special checking list, which will ensure safety in routine operation and even in abnormality. (author). 8 refs.

  16. Conceptual design report for the ICPP spent nuclear fuel dry storage project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

  17. Safe conditioning of waste for final disposal. Vitrification of spent used fuel elements; Sichere Konditionierung zur Endlagerung. Verglasung von abgebrannten Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Niessen, Stefan; Blanc, Eric [Areva GmbH, Erlangen (Germany)

    2016-08-15

    The strategy for disposal of spent nuclear fuel in Germany requires an interim storage over a longer period. The used fuel assemblies are stored in dry storage casks. An alternative method for storage is the conditioning of the fuel elements. This technology is proven on an industrial scale and is carried out at the La Hague plant. The know-how is currently available for both, the operators as well as in industry and science in Germany.

  18. Spent nuclear fuel storage. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The bibliography contains citations concerning spent nuclear fuel storage technologies, facilities, sites, and assessment. References review wet and dry storage, spent fuel casks and pools, underground storage, monitored and retrievable storage systems, and aluminum-clad spent fuels. Environmental impact, siting criteria, regulations, and risk assessment are also discussed. Computer codes and models for storage safety are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  19. Referenced-site environmental document for a Monitored Retrievable Storage facility: backup waste management option for handling 1800 MTU per year

    Energy Technology Data Exchange (ETDEWEB)

    Silviera, D.J.; Aaberg, R.L.; Cushing, C.E.; Marshall, A.; Scott, M.J.; Sewart, G.H.; Strenge, D.L.

    1985-06-01

    This environmental document includes a discussion of the purpose of a monitored retrievable storage facility, a description of two facility design concepts (sealed storage cask and field drywell), a description of three reference sites (arid, warm-wet, and cold-wet), and a discussion and comparison of the impacts associated with each of the six site/concept combinations. This analysis is based on a 15,000-MTU storage capacity and a throughput rate of up to 1800 MTU per year.

  20. Criticality safety evaluation report for the Cold Vacuum Drying Facility`s process water handling system

    Energy Technology Data Exchange (ETDEWEB)

    Roblyer, S.D.

    1998-02-12

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility (CVDF). The controls and limitations on equipment design and operations to control potential criticality occurrences are identified. The effectiveness of equipment design and operation controls in preventing criticality occurrences during normal and abnormal conditions is evaluated and documented in this report. Spent nuclear fuel (SNF) is removed from existing canisters in both the K East and K West Basins and loaded into a multicanister overpack (MCO) in the K Basin pool. The MCO is housed in a shipping cask surrounded by clean water in the annulus between the exterior of the MCO and the interior of the shipping cask. The fuel consists of spent N Reactor and some single pass reactor fuel. The MCO is transported to the CVDF near the K Basins to remove process water from the MCO interior and from the shipping cask annulus. After the bulk water is removed from the MCO, any remaining free liquid is removed by drawing a vacuum on the MCO`s interior. After cold vacuum drying is completed, the MCO is filled with an inert cover gas, the lid is replaced on the shipping cask, and the MCO is transported to the Canister Storage Building. The process water removed from the MCO contains fissionable materials from metallic uranium corrosion. The process water from the MCO is first collected in a geometrically safe process water conditioning receiver tank. The process water in the process water conditioning receiver tank is tested, then filtered, demineralized, and collected in the storage tank. The process water is finally removed from the storage tank and transported from the CVDF by truck.

  1. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  2. Draft report: Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions.

  3. Physics Flash August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kippen, Karen Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-25

    Physics Flash is the newsletter for the Physics Division at Los Alamos National Laboratory. This newsletter is for August 2016. The following topics are covered: "Accomplishments in the Trident Laser Facility", "David Meyerhofer elected as chair-elect APS Nominating Committee", "HAWC searches for gamma rays from dark matter", "Proton Radiography Facility commissions electromagnetic magnifier", and "Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks."

  4. Spent fuel management of Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

    Blanco Zurro, J.E.; Garcia Costilla, M. [Area de Generacion - Unidad Nuclear, Gas Natural Fenosa, Avda. de San Luis, 77, 28033 Madrid (Spain); Lavara Sanz, A. [Division Nuclear, SOCOIN, P. del Club Deportivo, 1 - Edificio 5, 28223 Pozuelo de Alarcon, Madrid (Spain); Martinez Abad, J.E. [Departamento de Residuos de Alta Actividad, ENRESA, C/ Emilio Vargas, 7, 28043 Madrid (Spain)

    2010-07-01

    The definitive shutdown of Jose Cabrera Nuclear Power Plant took place on 30. of April 2006. From this moment, cooperation agreements between ENRESA and GAS NATURAL FENOSA were established to reach, among others objectives, its decommissioning, 3 years after the shutdown of the reactor. In order to accomplish the Spanish nuclear regulation, a spent fuel management plan was developed. This plan determined that the fuel assemblies placed in the spent fuel pool would be managed by means of their storage in an interim installation. For this reason, an Independent Spent Fuel Storage Installation (ISFSI) was built at plant site, pioneer in Spain by its characteristics of design. Different administrative authorizations from the point of view of nuclear safety as well as from the environmental were required for ISFSI licensing process. The transference and storage of spent fuel was carried out using the HI-STORM 100Z Dry Storage System, developed by HOLTEC INTERNATIONAL. This system, designed for the spent fuel storage in casks, supports abnormal and very hard accident conditions. The system has three main components: Storage Cask (HI-STORM), Transfer Cask (HI-TRAC) and Multipurpose Canister (MPC). In addition to this, the system has a specific Transport Cask (HI-STAR) for the future transport out of the Plant. More than 30 Design Modifications to the system and plant were implemented to solve structural problems and to include safety and ALARA improvements. The transfer of the spent fuel and its emplacement in the ISFSI began on January 2009 and finished on September of that year allowing starting the decommissioning process, three years and a half after Jose Cabrera NPP shutdown. (authors)

  5. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  6. Nuclear power development and nuclear data activities in Malaysia

    Energy Technology Data Exchange (ETDEWEB)

    Gui Ah Auu [Malaysian Institute for Nuclear Technology Research, Ministry of Science, Technology and the Environment, Selangor (Malaysia)

    1999-03-01

    In this paper, research activities on nuclear power requirement carried out jointly by MINT and other organizations are described. Also discussed are activities on neutronics such as TRIGA reactor fuel management, storage pool criticality, and reactor fuel transfer cask calculations. In addition, recent work on radiation transport activities in MINT such as skyshine and photon phantom dose calculations using the MCNP and MRIPP computer codes are presented. Finally, nuclear data measurement works by researchers in Malaysian universities are described. (author)

  7. 327 SNF fuel return to K-Basin quality process plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-09-22

    The B and W Hanford Company`s (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE.

  8. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

  9. Byproducts Utilization Program: Sewage Sludge Irradiation Project. Progress report, July-December 1983

    Energy Technology Data Exchange (ETDEWEB)

    1984-12-01

    Engineering support for a demonstration-scale irradiator design included assisting the City of Albuquerque in preparing a comprehensive site plan for their proposed sludge handling and treatment facilities. The solar sludge dryer has been delivered to SNLA. A preliminary sludge drying experiment indicated the importance of optimizing stirring and air flow. Installation of instrumentation and mechanical equipment continued. The Sandia Irradiator for Dried Sewage Solids (SIDSS) was used to irradiate 23 tons of dried, digested sewage sludge for the New Mexico State University (NMSU) Department of Crop and Soil Sciences. Gamma Irradiation Facility (GIF) operations included irradiation of ground pork for Toxoplasma gondii inactivation experiments, irradiation of surgical supplies and soil samples. Beneficial Uses Shipping Systems (BUSS) cask activities included near completion of the two full-scale cask bodies. Work continued on the Cask Safety Analysis Report (SAR) including additional analyses to reconfigure the six strontium fluoride capsules and/or reduce the number of capsules accommodated. NMSU has indicated no regrowth of salmonellae occurred in the irradiated sludge stockpile, while salmonellae did regrow in the unirradiated stockpile. Analyses of raw and digested sewage sludge from the Albuquerque Waste Water Treatment Plant showed levels of Yersinia enterocolitica (a human pathogen of emerging significance) to be below detection limits.

  10. Studies on the production and application of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hyon Soo; Park, K. B.; Kim, J. R.; Yoon, B. M.; Bang, H. S.; Shin, B. C.; Cho, W. K.; Park, U. J.; Park, C. D.; Lee, Y. G.; Suh, C. H.; Shin, H. Y.; Kim, D. S.; Hong, S. B.; Jun, S. S.; Min, E. S.; Jang, K. D.; Kim, J. K.; Kim, S. J.; Yang, S. Y.; Yang, S. H.; Chun, K. J.; Kang, H. Y.; Suh, K. S.; Goo, J. H.; Chung, S. H.; Lee, J. C.; Choi, J. L.; Lee, H. Y.; Bang, K. S.

    1997-09-01

    To produce radioisotopes utilizing the research reactor `HANARO`, development of RI production process, target fabrication, preparation of devices and tools for RI process, preparation of production facility for radiopharmaceuticals, test production for the established process, etc. have been carried out, respectively. Production processes for various kinds of radionuclides were developed and the settled methods were applied to test production using `HANARO`. The results of developed process are as follows: (1) I-131 dry distillation method. (2) Large scale production of Ir-192 sources (3) P-32 production process by distillation under reduced pressure (4) Cr-51 production process using enriched target. To irradiate the target for RI production in `HANARO`, target for neutron irradiation, loading/unloading devices, working table in service pool, remote handling tools, shield cask for irradiated target transfer, etc. were designed and fabricated. The function test of prepared targets and the safety analysis of shielding casks were carried out. License for practical use of the prepared casks were obtained from Ministry of Science and Technology. For production of medical radioisotopes, their production facilities were designed in detail and were installed in RIPF (Radioisotope Production Facility), with full reflection of the basic concept of the good manufacturing practice for radiopharmaceuticals. The constructed GMP facilities have started to be operated after authorization since Jun., 1997. Results of this study will be applied to mass production of radioisotopes in `HANARO` and are to contribute the advance of domestic medicine and industry related to radioisotopes. (author). 7 refs., 7 tabs., 4 figs.

  11. A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.; Sindelar, R.; Duncan, A.; Adams, T.

    2014-04-07

    A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

  12. 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.

  13. DURABILITY OF DEPLETED URANIUM AGGREGATES (DUAGG) IN DUCRETE SHIELDING APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, Catherine H.; Dole, Leslie R.

    2003-02-27

    The depleted uranium (DU) inventory in the United States exceeds 500,000 metric tonnes. To evaluate the possibilities for reuse of this stockpile of DU, the U.S. Department of Energy (DOE) has created a research and development program to address the disposition of its DU(1). One potential use for this stockpile material is in the fabrication of nuclear shielding casks for the storage, transport, and disposal of spent nuclear fuels. The use of the DU-based shielding would reduce the size and weight of the casks while allowing a level of protection from neutrons and gamma rays comparable to that afforded by steel and concrete. DUAGG (depleted uranium aggregate) is formed of depleted uranium dioxide (DUO2) sintered with a synthetic-basalt-based binder. This study was designed to investigate possible deleterious reactions that could occur between the cement paste and the DUAGG. After 13 months of exposure to a cement pore solution, no deleterious expansive mineral phases were observed to form either with the DUO2 or with the simulated-basalt sintering phases. In the early stages of these exposure tests, Oak Ridge National Laboratory preliminary results confirm that the surface reactions of this aggregate proceed more slowly than expected. This finding may indicate that DUAGG/DUCRETE (depleted uranium concrete) casks could have service lives sufficient to meet the projected needs of DOE and the commercial nuclear power industry.

  14. Used fuel storage monitoring using novel 4He scintillation fast neutron detectors and neutron energy discrimination analysis

    Science.gov (United States)

    Kelley, Ryan P.

    With an increasing quantity of spent nuclear fuel being stored at power plants across the United States, the demand exists for a new method of cask monitoring. Certifying these casks for transportation and long-term storage is a unique dilemma: their sealed nature lends added security, but at the cost of requiring non-invasive measurement techniques to verify their contents. This research will design and develop a new method of passively scanning spent fuel casks using 4He scintillation detectors to make this process more accurate. 4He detectors are a relatively new technological development whose full capabilities have not yet been exploited. These detectors take advantage of the high 4He cross section for elastic scattering at fast neutron energies, particularly the resonance around 1 MeV. If one of these elastic scattering interactions occurs within the detector, the 4He nucleus takes energy from the incident neutron, then de-excites by scintillation. Photomultiplier Tubes (PMTs) at either end of the detector tube convert this emitted light into an electrical signal. The goal of this research is to use the neutron spectroscopy features of 4He scintillation detectors to maintain accountability of spent fuel in storage. This project will support spent fuel safeguards and the detection of fissile material, in order to minimize the risk of nuclear proliferation and terrorism.

  15. Moving hot cell for LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1994-09-16

    A moving hot cell for an LMFBR type reactor is made movable on a reactor operation floor between a position just above the reactor container and a position retreated therefrom. Further, it comprises an overhung portion which can incorporate a spent fuel just thereunder, and a crane for moving a fuel assembly between a spent fuel cask and a reactor container. Further, an opening/closing means having a shielding structure is disposed to the bottom portion and the overhung portion thereof, to provide a sealing structure, in which only the receiving port for the spent fuel cask faces to the inner side, and the cask itself is disposed at the outside. Upon exchange of fuels, the movable hot cell is placed just above the reactor to take out the spent fuels, so that a region contaminated with primary sodium is limited within the hot cell. On the other hand, upon maintenance and repair for equipments, the hot cell is moved, thereby enabling to provide a not contaminated reactor operation floor. (N.H.).

  16. USED FUEL RAIL SHOCK AND VIBRATION TESTING OPTIONS ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Best, Ralph E.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Maheras, Steven J.

    2014-09-29

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data that are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges

  17. Multi Canister Overpack (MCO) Design Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    GOLDMANN, L.H.

    2000-02-29

    The MCO is designed to facilitate the removal, processing and storage of the spent nuclear fuel currently stored in the East and West K-Basins. The MCO is a stainless steel canister approximately 24 inches in diameter and 166 inches long with cover cap installed. The shell and the collar which is welded to the shell are fabricated from 304/304L dual certified stainless steel for the shell and F304/F304L dual certified for the collar. The shell has a nominal thickness of 1/2 inch. The top closure consists of a shield plug with four processing ports and a locking ring with jacking bolts to pre-load a metal seal under the shield plug. The fuel is placed in one of four types of baskets, excluding the SPR fuel baskets, in the fuel retention basin. Each basket is then loaded into the MCO which is inside the transfer cask. Once all of the baskets are loaded into the MCO, the shield plug with a process tube is placed into the open end of the MCO. This shield plug provides shielding for workers when the transfer cask, containing the MCO, is lifted from the pool. After being removed from the pool, the locking ring is installed and the jacking bolts are tightened to pre-load the metal main closure seal. The cask is then sealed and the MCO taken to the Cold Vacuum Drying (CVD) facility for bulk water removal and vacuum drying through the process ports. Covers for the process ports may be installed or removed as needed per operating procedures. The MCO is then transferred to the Canister Storage Building (CSB), in the closed transfer cask. At the CSB, the MCO is then removed from the cask and becomes one of two MCOs stacked in a storage tube. MCOs will have a cover cap welded over the shield plug providing a complete welded closure. A number of MCOs may be stored with just the mechanical seal to allow monitoring of the MCO pressure, temperature, and gas composition.

  18. Intense Laser - Electron Beam Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, T.; Ditmire, T.; LeSage, G.

    2000-02-25

    Applicants seeking a Certificate of Compliance for an Independent Spent Fuel Storage Installation (ISFSI) cask must evaluate the consequences of a handling accident resulting in a drop or tip-over of the cask onto a concrete storage pad. As a result, analytical modeling approaches that might be used to evaluate the impact of cylindrical containers onto concrete pads are needed. One such approach, described and benchmarked in NUREG/CR-6608,{sup 1} consists of a dynamic finite element analysis using a concrete material model available in DYNA3D{sup 2} and in LS-DYNA,{sup 3} together with a method for post-processing the analysis results to calculate the deceleration of a solid steel billet when subjected to a drop or tip-over onto a concrete storage pad. The analysis approach described in NUREG/CR-6608 gives a good correlation of analysis and test results. The material model used for the concrete in the analyses in NUREG/CR-6608 is, however, somewhat troublesome to use, requiring a number of material constants which are difficult to obtain. Because of this a simpler approach, which adequately evaluates the impact of cylindrical containers onto concrete pads, is sought. Since finite element modeling of metals, and in particular carbon and stainless steel, is routinely and accurately accomplished with a number of finite element codes, the current task involves a literature search for and a discussion of available concrete models used in finite element codes. The goal is to find a balance between a concrete material model with a limited number of required material parameters which are readily obtainable, and a more complex model which is capable of accurately representing the complex behavior of the concrete storage pad under impact conditions. The purpose of this effort is to find the simplest possible way to analytically represent the storage cask deceleration during a cask tip-over or a cask drop onto a concrete storage pad. This report is divided into three sections

  19. Used fuel rail shock and vibration testing options analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, Nicholas A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-25

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data that are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges

  20. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  1. Minimisation of liquid radioactive operational wastes from light water reactors; Minimierung fluessiger radioaktiver Betriebsabfaelle aus Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Krumpholz, Udo [Kernkraftwerk Gundremmingen GmbH, Gundremmingen (Germany). Teilbereich Ueberwachung Chemie / Entsorgung, PNG-UC

    2014-12-15

    A system for decontaminating evaporator concentrates has been developed during R and D work at the Gundremmingen (KGG) nuclear power plant, by means of which accumulation of radioactive wastes can be effectively reduced. A cooling crystallization system is involved in this case, which extracts the high percentage of non-radioactive salt components from the brines through these salts being crystallised with a high level of purity and thereby being withdrawn from the nuclear disposal procedure. A method is also available in modified form for decontaminating concentrates containing boron from PWR plants. Use of cooling crystallisation renders superfluous the otherwise usual stages of waste treatment such as for example disposal scheduling, provision of repository casks (e.g. MOSAIK {sup registered}), their transport, packing, compilation of waste package documentation, intermediate storage and final disposal. Disposal of evaporator concentrates has no longer been necessary in KGG since 1998. It has been possible to avoid more than 500 MOSAIK {sup registered} type II casks in KGG since the procedure has been employed. Owing to the current price basis, a saving on the order of >30 million Euro has been achieved merely for cask acquisition since the procedure has been used. In addition to these advantages, operation of the cooling crystallisation system (KKA) is also reflected in a considerable dose re-duction for the personnel performing the operations, thereby fulfilling the objective derived from the German radiation protection ordinance (StrlSchV) of dose minimisation (avoidance of unnecessary exposure to radiation and dose reduction, paragraph 6 StrlSchV). Internatonal trade mark rights exist for the cooling crystallisation and boric acid decontamination procedure.

  2. Depleted uranium hexafluoride: The source material for advanced shielding systems

    Energy Technology Data Exchange (ETDEWEB)

    Quapp, W.J.; Lessing, P.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Cooley, C.R. [Department of Technology, Germantown, MD (United States)

    1997-02-01

    The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

  3. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilke, Jason [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Margraf, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  4. Changes in tension regulates proliferation and migration of fibroblasts by remodeling expression of ECM proteins

    Science.gov (United States)

    Jiang, Minmin; Qiu, Juhui; Zhang, Lingling; Lü, Dongyuan; Long, Mian; Chen, Li; Luo, Xiangdong

    2016-01-01

    Wound healing is a complicated but highly organized process in which cell migration and proliferation are actively involved. However, the process by which mechanical stretch regulates the proliferation and migration of human skin fibroblasts (HFs) and keratinocytes is poorly understood. Using a house built mechanical stretch device, we examined the HFs extracellular matrix (ECM) components changes under non-stretch, static stretch or cyclic stretch conditions. We further investigated the changes in ECM component protein expression levels in keratinocytes and analyzed the effects of individual ECM component on keratinocyte proliferation and migration. Particularly, the roles of calcium/calmodulin-dependent serine protein kinase (CASK) in the HF proliferation under cyclic stretch were investigated. Cyclic stretch suppressed HF proliferation compared with HFs without stretch or with static stretch. Cyclic stretch also led to a significant reduction in the levels of collagen I and a marked increase of fibronectin in HFs ECM. By contrast, collagen I levels increased and fibronectin levels decreased in response to non-stretch and static stretch conditions. After cyclic stretch, the proliferation of keratinocytes was inhibited by the cyclic stretch-induced ECM in HFs. The inoculation of keratinocytes with single ECM component suggested that collagen I was more capable of inducing cell proliferation than fibronectin, while it had less impact on cell migration compared with fibronectin. Furthermore, cyclic stretch induced by proliferation inhibition was associated with altered integrin β1-CASK signal pathway. The present results demonstrated the existence of HF-ECM-keratinocyte ‘cross-talk’ in cutaneous tissues. Thus, the integrin β1-CASK signal pathway in HFs may be involved in the outside-in signal transduction of extracellular stretch and the altered ECM component expression. PMID:27588075

  5. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  6. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  7. SLUDGE TREATMENT PROJECT COST COMPARISON BETWEEN HYDRAULIC LOADING AND SMALL CANISTER LOADING CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    GEUTHER J; CONRAD EA; RHOADARMER D

    2009-08-24

    The Sludge Treatment Project (STP) is considering two different concepts for the retrieval, loading, transport and interim storage of the K Basin sludge. The two design concepts under consideration are: (1) Hydraulic Loading Concept - In the hydraulic loading concept, the sludge is retrieved from the Engineered Containers directly into the Sludge Transport and Storage Container (STSC) while located in the STS cask in the modified KW Basin Annex. The sludge is loaded via a series of transfer, settle, decant, and filtration return steps until the STSC sludge transportation limits are met. The STSC is then transported to T Plant and placed in storage arrays in the T Plant canyon cells for interim storage. (2) Small Canister Concept - In the small canister concept, the sludge is transferred from the Engineered Containers (ECs) into a settling vessel. After settling and decanting, the sludge is loaded underwater into small canisters. The small canisters are then transferred to the existing Fuel Transport System (FTS) where they are loaded underwater into the FTS Shielded Transfer Cask (STC). The STC is raised from the basin and placed into the Cask Transfer Overpack (CTO), loaded onto the trailer in the KW Basin Annex for transport to T Plant. At T Plant, the CTO is removed from the transport trailer and placed on the canyon deck. The CTO and STC are opened and the small canisters are removed using the canyon crane and placed into an STSC. The STSC is closed, and placed in storage arrays in the T Plant canyon cells for interim storage. The purpose of the cost estimate is to provide a comparison of the two concepts described.

  8. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    Energy Technology Data Exchange (ETDEWEB)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  9. Development of a brittle fracture acceptance criterion for the International Atomic Energy Agency (IAEA)

    Energy Technology Data Exchange (ETDEWEB)

    Sorenson, K.B.; Salzbrenner, R. (Sandia National Labs., Albuquerque NM (United States)); Nickell, R.E.

    1993-01-01

    An effort has been undertaken to develop a brittle fracture acceptance criterion for structural components of nuclear material transportation casks. The need for such a criterion was twofold. First, new generation cask designs have proposed the use of ferritic steels and other materials to replace the austenitic stainless steel commonly used for structural components in transport casks. Unlike austenitic stainless steel which fails in a high-energy absorbing, ductile tearing mode, it is possible for these candidate materials to fail via brittle fracture when subjected to certain combinations of elevated loading rates and low temperatures. Second, there is no established brittle fracture criterion accepted by the regulatory community that covers a broad range of structural materials. Although the existing IAEA Safety Series no.37 addressed brittle fracture, its the guidance was dated and pertained only to ferritic steels. Consultant's Services Meetings held under the auspices of the IAEA have resulted in a recommended brittle fracture criterion. The brittle fracture criterion is based on linear elastic fracture mechanics, and is the result of a consensus of experts from six participating IAEA-member countries. The brittle fracture criterion allows three approaches to determine the fracture toughness of the structural material. The three approaches present the opportunity to balance material testing requirements and the conservatism of the material's fracture toughness which must be used to demonstrate resistance to brittle fracture. This work has resulted in a revised Appendix IX to Safety Series no.37 which will be released as an IAEA Technical Document within the coming year. (author).

  10. Development of a brittle fracture acceptance criterion for the International Atomic Energy Agency (IAEA)

    Energy Technology Data Exchange (ETDEWEB)

    Sorenson, Ken B.; Salzbrenner, Richard [Sandia National Labs., Albuquerque, NM (United States); Nickell, Robert E. [Applied Science and Technology, Inc., Poway, CA (United States)

    1992-01-01

    An effort has been undertaken to develop a brittle fracture acceptance criterion for structural components of nuclear material transportation casks. The need for such a criterion was twofold. First, new generation cask designs have proposed the use of ferritic steels and other materials to replace the austenitic stainless steel commonly used for structural components in transport casks. Unlike austenitic stainless steel which fails in a high-energy absorbing, ductile tearing mode, it is possible for these candidate materials to fail via brittle fracture when subjected to certain combinations of elevated loading rates and low temperatures. Second, there is no established brittle fracture criterion accepted by the regulatory community that covers a broad range of structural materials. Although the existing IAEA Safety Series {number sign}37 addressed brittle fracture, its the guidance was dated and pertained only to ferritic steels. Consultant's Services Meetings held under the auspices of the IAEA have resulted in a recommended brittle fracture criterion. The brittle fracture criterion is based on linear elastic fracture mechanics, and is the result of a consensus of experts from six participating IAEA-member countries. The brittle fracture criterion allows three approaches to determine the fracture toughness of the structural material. The three approaches present the opportunity to balance material testing requirements and the conservatism of the material's fracture toughness which must be used to demonstrate resistance to brittle fracture. This work has resulted in a revised Appendix IX to Safety Series {number sign}37 which will be released as an IAEA Technical Document within the coming year.

  11. STP-ECRTS - THERMAL AND GAS ANALYSES FOR SLUDGE TRANSPORT AND STORAGE CONTAINER (STSC) STORAGE AT T PLANT

    Energy Technology Data Exchange (ETDEWEB)

    CROWE RD; APTHORPE R; LEE SJ; PLYS MG

    2010-04-29

    The Sludge Treatment Project (STP) is responsible for the disposition of sludge contained in the six engineered containers and Settler tank within the 105-K West (KW) Basin. The STP is retrieving and transferring sludge from the Settler tank into engineered container SCS-CON-230. Then, the STP will retrieve and transfer sludge from the six engineered containers in the KW Basin directly into a Sludge Transport and Storage Containers (STSC) contained in a Sludge Transport System (STS) cask. The STSC/STS cask will be transported to T Plant for interim storage of the STSC. The STS cask will be loaded with an empty STSC and returned to the KW Basin for loading of additional sludge for transportation and interim storage at T Plant. CH2MHILL Plateau Remediation Company (CHPRC) contracted with Fauske & Associates, LLC (FAI) to perform thermal and gas generation analyses for interim storage of STP sludge in the Sludge Transport and Storage Container (STSCs) at T Plant. The sludge types considered are settler sludge and sludge originating from the floor of the KW Basin and stored in containers 210 and 220, which are bounding compositions. The conditions specified by CHPRC for analysis are provided in Section 5. The FAI report (FAI/10-83, Thermal and Gas Analyses for a Sludge Transport and Storage Container (STSC) at T Plant) (refer to Attachment 1) documents the analyses. The process considered was passive, interim storage of sludge in various cells at T Plant. The FATE{trademark} code is used for the calculation. The results are shown in terms of the peak sludge temperature and hydrogen concentrations in the STSC and the T Plant cell. In particular, the concerns addressed were the thermal stability of the sludge and the potential for flammable gas mixtures. This work was performed with preliminary design information and a preliminary software configuration.

  12. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  13. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-11-18

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  14. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  15. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup.

  16. The morphology, topography and cytoarchitectonics of the ciliary ganglion in the domestic turkey (Meleagris gallopavo domesticus).

    Science.gov (United States)

    Radzimirska, Małgorzata

    2003-11-01

    The ciliary ganglion of the domestic turkey (Meleagris gallopavo domesticus) is located between the posterior wall of the eyeball and the optic nerve. It is closely connected with the oculomotor nerve; in particular with its inferior branch. The ganglion has a cask-like shape and is adjacent to the inferior branch of the oculomotor nerve. From this ganglion postganglionic fibres emerge which are arranged in two fasciculi. These are termed the long ciliary nerves and the short ciliary nerves. A cross-section of the ciliary ganglion revealed two populations of cells: small ones - choroid cells and large ones - ciliary cells.

  17. Report on UQ and PCMM Analysis of Vacuum Drying for UFD S&T Gaps

    Energy Technology Data Exchange (ETDEWEB)

    Fluss, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-31

    This report discusses two phenomena that could affect the safety, licensing, transportation, storage, and disposition of the spent fuel storage casks and their contents (radial hydriding during drying and water retention after drying) associated with the drying of canisters for dry spent fuel storage. The report discusses modeling frameworks and evaluations that are, or have been, developed as a means to better understand these phenomena. Where applicable, the report also discusses data needs and procedures for monitoring or evaluating the condition of storage containers during and after drying. A recommendation for the manufacturing of a fully passivated fuel rod, resistant to oxidation and hydriding is outlined.

  18. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  19. Extending dry storage of spent LWR fuel for 100 years.

    Energy Technology Data Exchange (ETDEWEB)

    Einziger, R. E.

    1998-12-16

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  20. Development of dry storage technology of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Maruoka, Kunio [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Nuclear Energy Systems Engineering Center; Murakami, Kazuo; Yokoyama, Takeshi; Natsume, Tomohiro; Irino, Mitsuhiro

    1998-07-01

    The increasing demand for storage of spent fuel assemblies generated by commercial nuclear power plants is the urgent subject to solve. The dry storage system is as economically more advantageous than the pool storage system, and so, Mitsubishi Heavy Industries, Ltd. has developed the metal storage cask suited to small and medium storage capacity under 2000MTU - 3000MTU. For large scale capacity, the new `Mitsubishi Vault Storage System` has been developed, and it provides a safe and economical solution. Technical study concerning cooling ability was performed. (author)

  1. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    D.T. Dexheimer

    2004-02-27

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Canister Handling Facility (CHF) performing operations to receive transportation casks, transfer wastes, prepare waste packages, perform associated equipment maintenance. The specific scope of work contained in this calculation covers individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results of this calculation will be used to support the design of the CHF and provide occupational dose estimates for the License Application.

  2. Safety evaluation for packaging (onsite) for cesium chloride capsules with type W overpacks

    Energy Technology Data Exchange (ETDEWEB)

    McCoy, J.C.

    1997-09-15

    This Safety Evaluation for Packaging (SEP) documents the evaluation of a new basket design and overpacked cesium chloride capsule payload for the Beneficial Uses Shipping System (BUSS) Cask in accordance with the onsite transportation requirements of the Hazardous Material Packaging and Shipping manual, WHC-CM-2-14. This design supports the one-time onsite shipment of 16 cesium chloride capsules with Type W overpacks from the 324 Building to the 224T Building at the Waste Encapsulation and Storage Facility (WESF). The SEP is valid for a one-time onsite shipment or until August 1, 1998, whichever occurs first.

  3. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Lee, H. H.; Kim, K. H. and others

    2000-03-01

    The objectives of this study are (1) the refurbishment for PIEF(Post Irradiation Examination Facility) and M6 hot-cell in IMEF(Irradiated Material Examination Facility), (2) the establishment of the compatible facility for DUPIC fuel fabrication experiments which is licensed by government organization, and (3) the establishment of the transportation system and transportation cask for nuclear material between facilities. The report for this project describes following contents, such as objectives, necessities, scope, contents, results of current step, R and D plan in future and etc.

  4. Dry Transfer Systems for Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  5. ENSA: Engineering and Design; ENSA: Ingenieria y Diseno

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    Since 1973, ENSA has been present in the nuclear market as a supplier of primary components for newly built nuclear power plants and also replacement components for already existing plants, in accordance with the foundational purpose for which it was created. Simultaneously with the above, it has in recent years entered the services business and the field of design and manufacture of pool fuel storage racks, storage and transportation casks, non-destructive testing (NDT) validation models, etc. More recently, it has participated significantly in projects involving the so-called Generation IV reactors, such as the PBMR in South Africa, the HTR in China and the Westinghouse-sponsored IRIS. (Author)

  6. Development of boron carbide-copper cermets. Status report

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    The status of a program to develop a B/sub 4/C-Cu cermet for Breeder Reactor spent-fuel shipping cask neutron shields is presented. It is shown that inspectable 6 to 7 cm thick 60 to 70 volume percent B/sub 4/C cermets can be fabricated using hot isostatic powder processing procedures. An alternative manufacturing method, rheocasting, also appears to be a promising, perhaps more cost-effective method for producing these cermets. Recommendations for further development of these manufacturing processes are given.

  7. Savannah River Technology Center monthly report: June 1994

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    Short summaries are given for 53 projects concerned with tritium, separations, environmental, waste management and general topics. Included in the general topics are the following: High-level waste integrated flowsheet model; Scoping thermal analysis of failed equipment storage vault thermal model; Decision support system for in-tank precipitation benzene stripper experiment; Development of a liquid level bubbler for Defense Waste Processing Facility canyon vessels; Thermal testing of 9973 and 9975 series packagings; and Technical review of CD-1 cask for first time use.

  8. Existing Condition Analysis of Dry Spent Fuel Storage Technology

    Institute of Scientific and Technical Information of China (English)

    LI Ning; XU Lan; HAO Jian-sheng

    2016-01-01

    As in some domestic nuclear power plants, spent fuel pools near capacity, away-from-reactor type storage should be arranged to transfer spent fuel before the pool capacity is full and the plants can operate in safety. This study compares the features of wet and dry storage technology, analyzes the actualities of foreign dry storage facilities and then introduces structural characteristics of some foreign dry storage cask. Finally, a glance will be cast on the failure of away-from-reactor storage facilities of Pressurized Water Reactor(PWR)to meet the need of spent-fuel storage. Therefore, this study believes dry storage will be a feasible solution to the problem.

  9. 系统性红斑狼疮中内皮细胞相关自身抗原的鉴定%Identification of autoantigens associated with systemic lupus erythematosus from endothelial cells

    Institute of Scientific and Technical Information of China (English)

    曹毅; 吴乔; 牟芝蓉

    2012-01-01

    We aimed to identify new potential autoantigens associated with systemic lupus erythematosus (SLE) in human umbilical vein endothelial cells (HUVEC). Serum samples were collected from 27 SLE patients and 24 health donors. And IgG from serum was captured by Protein G Beads. Then the beads were washed and incubated with HUVEC lysate. The proteins which reacted with IgG-Beads were separated by immobilized pH gradient two -dimensional (2 -D) gel electrophoresis, and the unique protein spots in SLE patients group were identified by LC -MS -MS. We identified several proteins including known SLE autoantigen GAPDH, in which calcium/calmodulin-dependent serine protein kinase (CASK) was the first find to be involved in SLE. Furthermore, we also found that CASK antibody expression in SLE was significantly higher than that in health donor. From all the results, we presumed that CASK in HUVEC may be a new potential autoantigen in SLE. Systemic lupus erythematosus; Human umbilical vein endothelial cells; Calcium/calmodulin-%目的 在内皮细胞中筛选与系统性红斑狼疮(systemic lupus erythematosus,SLE)疾病相关的自身抗原.方法 用免疫沉淀法以SLE病人血清中自身抗体捕获人脐带内皮细胞(human umbilical vein endothelial cell,HUVEC)相关抗原,并用双向电泳法分离免疫沉淀产物,然后用LC-MS-MS串联质谱鉴定与SLE病人血清反应的蛋白点.最后用Western blot法验证部分鉴定蛋白.结果 相对于正常人血清对照,SLE病人血清捕获了多个内皮细胞相关蛋白,质谱鉴定结果 显示:通过免疫沉淀与双向电泳结合的方法 成功鉴定了包括GAPDH等已知SLE自身抗原在内的一系列蛋白,其中钙/钙调蛋白依赖性丝氨酸蛋白激酶(calcium/calmodulin-dependent serine protein kinase,CASK)为新发现SLE候选自身抗原.并以重组人CASK蛋白证实SLE病人血清中CASK抗体水平显著高于正常人对照.结论 内皮细胞蛋白CASK可能作为一个新的SLE相关自身抗原.

  10. Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

    Energy Technology Data Exchange (ETDEWEB)

    Annette L. Schafer; Lloyd C. Brown; David C. Carathers; Boyd D. Christensen; James J. Dahl; Mark L. Miller; Cathy Ottinger Farnum; Steven Peterson; A. Jeffrey Sondrup; Peter V. Subaiya; Daniel M. Wachs; Ruth F. Weiner

    2013-11-01

    Environmental and health impacts are presented for activities associated with transient testing of nuclear fuel and material using two candidate test reactors. Transient testing involves irradiation of nuclear fuel or materials for short time-periods under high neutron flux rates. The transient testing process includes transportation of nuclear fuel or materials inside a robust shipping cask to a hot cell, removal from the shipping cask, pre-irradiation examination of the nuclear materials, assembly of an experiment assembly, transportation of the experiment assembly to the test reactor, irradiation in the test reactor, transport back to the hot cell, and post-irradiation examination of the nuclear fuel or material. The potential for environmental or health consequences during the transportation, examination, and irradiation actions are assessed for normal operations, off-normal (accident) scenarios, and transportation. Impacts to the environment (air, soil, and groundwater), are assessed during each phase of the transient testing process. This report documents the evaluation of potential consequences to the general public. This document supports the Environmental Assessment (EA) required by the U.S. National Environmental Policy Act (NEPA) (42 USC Subsection 4321 et seq.).

  11. SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets

    Energy Technology Data Exchange (ETDEWEB)

    Allender, Jeffrey S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bridges, Nicholas J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Loftin, Bradley M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dunsmuir, Michael D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-14

    Savannah River National Laboratory (SRNL) and Oak Ridge National Laboratory (ORNL) are developing plans for the recovery of rare and unique isotopes contained within heavy-actinide target assemblies, specifically the Mark-18A. Mark-18A assemblies were irradiated in Savannah River Site (SRS) reactors in the 1970s under extremely high neutron-flux conditions and produced, virtually, the world's supply of plutonium-244, an isotope of key importance to high-precision actinide measurement and other scientific and nonproliferation uses; and curium highly enriched in heavy isotopes (e.g., curium-246 and curium-248). In 2015 and 2016, SRNL is pursuing tasks that would reduce program risk and budget requirements, including further characterization of unprocessed targets; engineering studies for the use of the SRNL Shielded Cells Facility (SCF) for recovery; and development of onsite and offsite shipping methods including a replacement for the heavy (70 ton) cask previously used for onsite transfer of irradiated items at SRS. A status update is provided for the characterization, including modeling using the Monte Carlo N-Particle Transport Code (MCNP); direct non-destructive assay measurements; and cask design.

  12. Surrogate/spent fuel sabotage : aerosol ratio test program and Phase 2 test results.

    Energy Technology Data Exchange (ETDEWEB)

    Borek, Theodore Thaddeus III; Thompson, N. Slater (U.S. Department of Energy); Sorenson, Ken Bryce; Hibbs, R.S. (U.S. Department of Energy); Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Brochard, Didier (Institut de Radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany)

    2004-05-01

    A multinational test program is in progress to quantify the aerosol particulates produced when a high energy density device, HEDD, impacts surrogate material and actual spent fuel test rodlets. This program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments; the program also provides significant political benefits in international cooperation. We are quantifying the spent fuel ratio, SFR, the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are crucial for predicting radiological impacts. This document includes a thorough description of the test program, including the current, detailed test plan, concept and design, plus a description of all test components, and requirements for future components and related nuclear facility needs. It also serves as a program status report as of the end of FY 2003. All available test results, observations, and analyses - primarily for surrogate material Phase 2 tests using cerium oxide sintered ceramic pellets are included. This spent fuel sabotage - aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC, and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission.

  13. Spent fuel sabotage test program, characterization of aerosol dispersal : technical review and analysis supplement.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2009-07-01

    This project seeks to provide vital data required to assess the consequences of a terrorist attack on a spent fuel transportation cask. One such attack scenario involves the use of conical shaped charges (CSC), which are capable of damaging a spent fuel transportation cask. In the event of such an attack, the amount of radioactivity that may be released as respirable aerosols is not known with great certainty. Research to date has focused on measuring the aerosol release from single short surrogate fuel rodlets subjected to attack by a small CSC device in various aerosol chamber designs. The last series of three experiments tested surrogate fuel rodlets made with depleted uranium oxide ceramic pellets in a specially designed double chamber aerosol containment apparatus. This robust testing apparatus was designed to prevent any radioactive release and allow high level radioactive waste disposal of the entire apparatus following testing of actual spent fuel rodlets as proposed. DOE and Sandia reviews of the project to date identified a number of issues. The purpose of this supplemental report is to address and document the DOE review comments and to resolve the issues identified in the Sandia technical review.

  14. Conceptual design of the hot cell facility universal docking station at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Benchikhoune, M.; Friconneau, J.P.; Ivanov, V. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Lemee, A. [SOGETI High Tech, 180 Rue Rene Descartes, 13851 Aix en Provence (France); Martins, J.P. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Tamassy, G. [SOGETI High Tech, 180 Rue Rene Descartes, 13851 Aix en Provence (France)

    2011-10-15

    Between main shutdowns of the ITER machine, in-vessel components and Iter Remote Maintenance System (IRMS) are transferred between the Tokamak complex and the Hot Cell Facility using different types of sealed casks. Transfer Casks have different physical interfaces with the Vacuum Vessel, which need to be the same at the docking stations of the HCF. It means that in-vessel components and IRMS are cleaned in the same cells, which is in fact not convenient. Furthermore, logistic studies showed that the use rate of the cells is very inhomogeneous. In order to have dedicated cell for decontamination of Remote Handling tools, in order to increase the operability efficiency and to removes the hot cell docking operation from the critical path, the concept of a universal docking station has been investigated. Based on an existing design, the work was focused on a review of requirements, the re-design and the integration within the HCF layout. The universal docking station has been proposed and is now integrated in HCF design.

  15. Design of a high-level waste repository system for the United States

    Energy Technology Data Exchange (ETDEWEB)

    Baeza, J.L.; Boerigter, S.T.; Broadbent, G.E.; Cabello, E.D.; Duran, V.B.; Hollaway, W.R.; Karlberg, R.P.; Siegel, M.J.; Simonson, S.A.

    1988-05-12

    This report presents a conceptual design for a High Level Waste disposal system for fuel discharged by US commercial power reactors, using the Yucca Mountain repository site recently designated by federal legislation. Principal features of the resulting conceptual design include use of unit trains for periodic removal of old spent fuel from at-reactor storage facilities, buffer storage at the repository site using dual purpose transportation/storage casks, repackaging of the spent fuel from the dual purpose transportation/storage casks directly into special-alloy disposal canisters as intact fuel assemblies, without rod consolidation, emplacement into a repository of modular design, use of excavation techniques that minimize disturbance, both mechanical and chemical, to the geologic environment, a unit rail mounted vehicle for both the transportation and emplacement of the canister from the surface facilities to the underground repository, and a cost-effectiveness computer model of Yucca Mountain and an independent cost evaluation by members of the design team. 31 refs., 58 figs., 15 tabs.

  16. Develop an piezoelectric sensing based on SHM system for nuclear dry storage system

    Science.gov (United States)

    Ma, Linlin; Lin, Bin; Sun, Xiaoyi; Howden, Stephen; Yu, Lingyu

    2016-04-01

    In US, there are over 1482 dry cask storage system (DCSS) in use storing 57,807 fuel assemblies. Monitoring is necessary to determine and predict the degradation state of the systems and structures. Therefore, nondestructive monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health" for the safe operation of nuclear power plants (NPP) and radioactive waste storage systems (RWSS). Innovative approaches are desired to evaluate the degradation and damage of used fuel containers under extended storage. Structural health monitoring (SHM) is an emerging technology that uses in-situ sensory system to perform rapid nondestructive detection of structural damage as well as long-term integrity monitoring. It has been extensively studied in aerospace engineering over the past two decades. This paper presents the development of a SHM and damage detection methodology based on piezoelectric sensors technologies for steel canisters in nuclear dry cask storage system. Durability and survivability of piezoelectric sensors under temperature influence are first investigated in this work by evaluating sensor capacitance and electromechanical admittance. Toward damage detection, the PES are configured in pitch catch setup to transmit and receive guided waves in plate-like structures. When the inspected structure has damage such as a surface defect, the incident guided waves will be reflected or scattered resulting in changes in the wave measurements. Sparse array algorithm is developed and implemented using multiple sensors to image the structure. The sparse array algorithm is also evaluated at elevated temperature.

  17. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  18. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  19. Retrievable surface storage facility conceptual system design description

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    The studies evaluated several potentially attractive methods for processing and retrievably storing high-level radioactive waste after delivery to the Federal repository. These studies indicated that several systems could be engineered to safely store the waste, but that the simplest and most attractive concept from a technical standpoint would be to store the waste in a sealed stainless steel canister enclosed in a 2 in. thick carbon steel cask which in turn would be inserted into a reinforced concrete gamma-neutron shield, which would also provide the necessary air-cooling through an air annulus between the cask and the shield. This concept best satisfies the requirements for safety, long-term exposure to natural phenomena, low capital and operating costs, retrievability, amenability to incremental development, and acceptably small environmental impact. This document assumes that the reference site would be on ERDA's Hanford reservation. This document is a Conceptual System Design Description of the facilities which could satisfy all of the functional requirements within the established basic design criteria. The Retrievable Surface Storage Facility (RSSF) is planned with the capacity to process and store the waste received in either a calcine or glass/ceramic form. The RSSF planning is based on a modular development program in which the modular increments are constructed at rates matching projected waste receipts.

  20. Process and equipment development for hot isostatic pressing treatability study

    Energy Technology Data Exchange (ETDEWEB)

    Bateman, Ken; Wahlquist, Dennis; Malewitz, Tim

    2015-03-01

    Battelle Energy Alliance (BEA), LLC, has developed processes and equipment for a pilot-scale hot isostatic pressing (HIP) treatability study to stabilize and volume reduce radioactive calcine stored at Idaho National Laboratory (INL). In 2009, the U. S. Department of Energy signed a Record of Decision with the state of Idaho selecting HIP technology as the method to treat 5,800 yd^3 (4,400 m^3) of granular zirconia and alumina calcine produced between 1953 and 1992 as a waste byproduct of spent nuclear fuel reprocessing. Since the 1990s, a variety of radioactive and hazardous waste forms have been remotely treated using HIP within INL hot cells. To execute the remote process at INL, waste is loaded into a stainless-steel or aluminum can, which is evacuated, sealed, and placed into a HIP furnace. The HIP simultaneously heats and pressurizes the waste, reducing its volume and increasing its durability. Two 1 gal cans of calcine waste currently stored in a shielded cask were identified as candidate materials for a treatability study involving the HIP process. Equipment and materials for cask-handling and calcine transfer into INL hot cells, as well as remotely operated equipment for waste can opening, particle sizing, material blending, and HIP can loading have been designed and successfully tested. These results demonstrate BEA’s readiness for treatment of INL calcine.

  1. ASME Code requirements for multi-canister overpack design and fabrication

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1998-11-03

    The baseline requirements for the design and fabrication of the MCO include the application of the technical requirements of the ASME Code, Section III, Subsection NB for containment and Section III, Subsection NG for criticality control. ASME Code administrative requirements, which have not historically been applied at the Hanford site and which have not been required by the US Nuclear Regulatory Commission (NRC) for licensed spent fuel casks/canisters, were not invoked for the MCO. As a result of recommendations made from an ASME Code consultant in response to DNFSB staff concerns regarding ASME Code application, the SNF Project will be making the following modifications: issue an ASME Code Design Specification and Design Report, certified by a Registered Professional Engineer; Require the MCO fabricator to hold ASME Section III or Section VIII, Division 2 accreditation; and Use ASME Authorized Inspectors for MCO fabrication. Incorporation of these modifications will ensure that the MCO is designed and fabricated in accordance with the ASME Code. Code Stamping has not been a requirement at the Hanford site, nor for NRC licensed spent fuel casks/canisters, but will be considered if determined to be economically justified.

  2. The Global Threat Reduction Initiative's Return of Highly Enriched Uranium from Chile

    Energy Technology Data Exchange (ETDEWEB)

    Messick, C.E.; Dickerson, S.L.; Greenberg, R.F. Jr. [U.S. Department of Energy, National Nuclear Security Administration, Washington D.C. (United States); Andes, T.C. [Y-12 National Security Complex, Oak Ridge, TN (United States)

    2011-07-01

    In March 2010, the U.S. National Nuclear Security Administration's Office of Global Threat Reduction (GTRI), in collaboration with the Chilean Nuclear Energy Commission (CCHEN), completed a shipment of 18.2 kilograms of non-U.S.-origin highly enriched uranium (HEU) to the United States. The HEU was in the form of 71 aluminium-clad material test reactor (MTR) fuel elements and was the first GTRI Gap Program shipment that included non-U.S. origin irradiated nuclear fuel. Although shipments of research reactor fuels are not unique, this shipment served as a cornerstone to the first Presidential Nuclear Security Summit held in Washington, D.C., in April 2010. Carrying out the shipment became critical when a severe earthquake struck Chile just one day before the shipment was to occur. As the fuel had already been packaged in casks and the ocean vessels were nearing the port, U.S. and Chilean officials decided that it was most imperative that the shipment continue as planned. After careful analysis of the situation, inspection of the transportation packages, roadways, and port services, the shipment team was able to make the shipment occur in a safe and secure manner. This paper describes the loading activities at both the RECH-1 and RECH-2 reactors as well as the transportation of the loaded casks to the port of departure. (author)

  3. Inhibitory effects of whisky congeners on IgE-mediated degranulation in rat basophilic leukemia RBL-2H3 cells and passive cutaneous anaphylaxis reaction in mice.

    Science.gov (United States)

    Itoh, Tomohiro; Tsukane, Mariko; Koike, Minako; Nakamura, Chizu; Ohguchi, Kenji; Ito, Masafumi; Akao, Yukihiro; Koshimizu, Seiichi; Nozawa, Yoshinori; Wakimoto, Toshiyuki; Nukaya, Haruo; Suwa, Yoshihide

    2010-06-23

    Whisky is matured in oak casks. Many nonvolatile substances (whisky congeners, WC) seep from the oak cask during the maturing process. In this study, three antiallergic agents (syringaldehyde, SA; lyoniresinol, Lyo; and ellagic acid, EA) were isolated from WC. Treatment with SA, Lyo, and EA reduced the elevation of intracellular free Ca(2+) concentration ([Ca(2+)]i) and intracellular ROS production caused by FcepsilonRI activation. The inhibitions of the elevation of [Ca(2+)]i and intracellular ROS production by SA and Lyo were mainly due to the suppression of the NADPH oxidase activity and scavenging of the produced radical, respectively. On the other hand, EA inactivated spleen tyrosine kinase and led to the inhibition of the elevation of [Ca(2+)]i and intracellular ROS production. Furthermore, it was found that WC strongly inhibited IgE binding to the FcepsilonRIalpha chain, whereas SA, Lyo, and EA did not indicate this inhibitory effect. These results suggest that WC inhibits allergic reactions through multiple mechanisms. To disclose the in vivo effects of WC, SA, Lyo, and EA, these compounds were administered to type I allergic model mice, and the passive cutaneous anaphylaxis (PCA) reaction was measured. These compounds remarkably suppressed the PCA reaction. Taken together, these findings suggest that WC seemed to be beneficial to ameliorate allergic reactions.

  4. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Choi, J. W.; Go, W. I.; Kim, H. D.; Song, K. C.; Jeong, I. H.; Park, H. S.; Im, C. S.; Lee, H. M.; Moon, K. H.; Hong, K. P.; Lee, K. S.; Suh, K. S.; Kim, E. K.; Min, D. K.; Lee, J. C.; Chun, Y. B.; Paik, S. Y.; Lee, E. P.; Yoo, G. S.; Kim, Y. S.; Park, J. C.

    1997-09-01

    In the early stage of the project, a comprehensive survey was conducted to identify the feasibility of using available facilities and of interface between those facilities. It was found out that the shielded cell M6 interface between those facilities. It was found out that the shielded cell M6 of IMEF could be used for the main process experiments of DUPIC fuel fabrication in regard to space adequacy, material flow, equipment layout, etc. Based on such examination, a suitable adapter system for material transfer around the M6 cell was engineered. Regarding the PIEF facility, where spent PWR fuel assemblies are stored in an annex pool, disassembly devices in the pool are retrofitted and spent fuel rod cutting and shipping system to the IMEF are designed and built. For acquisition of casks for radioactive material transport between the facilities, some adaptive refurbishment was applied to the available cask (Padirac) based on extensive analysis on safety requirements. A mockup test facility was newly acquired for remote test of DUPIC fuel fabrication process equipment prior to installation in the M6 cell of the IMEF facility. (author). 157 refs., 57 tabs., 65 figs.

  5. WIPP Remote Handled Waste Facility: Performance Dry Run Operations

    Energy Technology Data Exchange (ETDEWEB)

    Burrington, T. P.; Britain, R. M.; Cassingham, S. T.

    2003-02-24

    The Remote Handled (RH) TRU Waste Handling Facility at the Waste Isolation Pilot Plant (WIPP) was recently upgraded and modified in preparation for handling and disposal of RH Transuranic (TRU) waste. This modification will allow processing of RH-TRU waste arriving at the WIPP site in two different types of shielded road casks, the RH-TRU 72B and the CNS 10-160B. Washington TRU Solutions (WTS), the WIPP Management and Operation Contractor (MOC), conducted a performance dry run (PDR), beginning August 19, 2002 and successfully completed it on August 24, 2002. The PDR demonstrated that the RHTRU waste handling system works as designed and demonstrated the handling process for each cask, including underground disposal. The purpose of the PDR was to develop and implement a plan that would define in general terms how the WIPP RH-TRU waste handling process would be conducted and evaluated. The PDR demonstrated WIPP operations and support activities required to dispose of RH-TRU waste in the WIPP underground.

  6. Criticality Analysis for Proposed Maximum Fuel Loading in a Standardized SNF Canister with Type 1a Baskets

    Energy Technology Data Exchange (ETDEWEB)

    Chad Pope; Larry L. Taylor; Soon Sam Kim

    2007-02-01

    This document represents a summary version of the criticality analysis done to support loading SNF in a Type 1a basket/standard canister combination. Specifically, this engineering design file (EDF) captures the information pertinent to the intact condition of four fuel types with different fissile loads and their calculated reactivities. These fuels are then degraded into various configurations inside a canister without the presence of significant moderation. The important aspect of this study is the portrayal of the fuel degradation and its effect on the reactivity of a single canister given the supposition there will be continued moderation exclusion from the canister. Subsequent analyses also investigate the most reactive ‘dry’ canister in a nine canister array inside a hypothetical transport cask, both dry and partial to complete flooding inside the transport cask. The analyses also includes a comparison of the most reactive configuration to other benchmarked fuels using a software package called TSUNAMI, which is part of the SCALE 5.0 suite of software.

  7. Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

    Energy Technology Data Exchange (ETDEWEB)

    LECHELT, J.A.

    2000-10-17

    The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix.

  8. Surrogate/spent fuel sabotage aerosol ratio testing:phase 1 summary and results.

    Energy Technology Data Exchange (ETDEWEB)

    Vigil, Manuel Gilbert; Sorenson, Ken Bryce; Lange, F. (Gesellschaft fur Anlagen- und reaktorsicherheit (GRS), Germany); Nolte, O. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Koch, W. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Dickey, Roy R.; Yoshimura, Richard Hiroyuki; Molecke, Martin Alan; Autrusson, Bruno (Institut de Radioprotection et de Surete Nucleaire (IRSN), France); Young, F. I. (U.S. Nuclear Regulatory Commission); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und reaktorsicherheit (GRS), Germany)

    2005-10-01

    This multinational test program is quantifying the aerosol particulates produced when a high energy density device (HEDD) impacts surrogate material and actual spent fuel test rodlets. The experimental work, performed in four consecutive test phases, has been in progress for several years. The overall program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This program also provides significant political benefits in international cooperation for nuclear security related evaluations. The spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC), and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission. This report summarizes the preliminary, Phase 1 work performed in 2001 and 2002 at Sandia National Laboratories and the Fraunhofer Institute, Germany, and documents the experimental results obtained, observations, and preliminary interpretations. Phase 1 testing included: performance quantifications of the HEDD devices; characterization of the HEDD or conical shaped charge (CSC) jet properties with multiple tests; refinement of the aerosol particle collection apparatus being used; and, CSC jet-aerosol tests using leaded glass plates and glass pellets, serving as representative brittle materials. Phase 1 testing was quite important for the design and performance of the following Phase 2 test program and test apparatus.

  9. Travel to France as Chief US Delegate at a meeting of International Standards Organization ISO/TC-85, ``Nuclear Technology``. Foreign trip report, March 17--March 26, 1994

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, R.M.

    1994-04-11

    As overall US Advisor for ISO/TC-85, SC-5, Dr. Westfall met with (1) Work Group 1, ``Measurement Techniques for the Chemical and Physical Characterization of UF{sub 6}, UO{sub 2}, and Mixed Oxide,`` on Monday, March 21, (2) Work Group 5, ``Standardization of Measurement Methods for the Characterization of Solid and Solidified Waste Forms, and for the Corrosion of their Primary Containers,`` on Tuesday, March 22; and (3) the full Subcommittee-5 on Wednesday, March 23. The status of work by all seven work groups in SC-5 was reported. Those having to do with nuclear fuel transportation (WG-4: UF, Containers, WG-9: Cask Trunnions, and WG-10: Cask Confinement) either have approved standards or drafts at an advanced stage of development. These work group convenors were asked to maintain their membership and establish new work areas in the field of nuclear fuel packaging. Definition of scope for new work is to be done in coordination with the interested staff members of the International Atomic Energy Agency, Vienna, Austria. On Thursday, March 24, the Cogema-Marcoule Plant staff hosted the SC-5 members to technical tours of their nuclear fuel reprocessing and waste vitrification and storage facilities.

  10. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

  11. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  12. UHMW-PE. A shielding material with special properties influenced by radiation

    Energy Technology Data Exchange (ETDEWEB)

    Ehe, K. von der; Jaunich, M.; Wolff, D. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2011-07-01

    Ultra high molecular weight polyethylene (UHMW-PE) - due to its high hydrogen content - is used as neutron shielding material in casks for storage and transport of radioactive materials. Besides this, UHMW-PE - as a high-performance polymer - has been used for several years in medical technology due to its excellent slip and wear properties. Due to the special properties of UHMW-PE which result basically from its extreme chain lengths and its high degree of crystallinity, it is predestinated for the aforementioned applications. In both cases, irradiation and its impact on the molecular structure of polyethylene play an important role. In the first case, irradiation exists as a by-product of inserting the radioactive material in the cask. Hence PE has to withstand any type of degradation affecting safety relevant aspects to be applicable for long term radiation shielding purposes for instance over a period of 40 years. In the second case irradiation is applied deliberately for purposes such as sterilization and crosslinking, leading to partial improvement of the mechanical properties (e.g. fracture toughness, crack propagation resistance, wear resistance) and better chemical stability. Specifically concerning their use in the field of medical technology, different types of UHMW-PE have been objects of numerous publications. It is generally accepted that two parallel and competitive processes, based on chain scission and reactions of C-centered radicals and molecular fragments, occur in PE as a consequence of radiation: radical recombination accounts for crosslinking, together with some disproportionation, formation of low molecular weight fragments, and recrystallization. Furthermore, formation of oxygenated structures in the presence of traces of oxygen is an antagonist of the C-centered radicals. Radiation induced scission preferentially takes place in the amorphous phase and noncrystalline surface of the crystals. It is followed by folding of molecular fragments

  13. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  14. Novel Nuclear Powered Photocatalytic Energy Conversion

    Energy Technology Data Exchange (ETDEWEB)

    White,John R.; Kinsmen,Douglas; Regan,Thomas M.; Bobek,Leo M.

    2005-08-29

    The University of Massachusetts Lowell Radiation Laboratory (UMLRL) is involved in a comprehensive project to investigate a unique radiation sensing and energy conversion technology with applications for in-situ monitoring of spent nuclear fuel (SNF) during cask transport and storage. The technology makes use of the gamma photons emitted from the SNF as an inherent power source for driving a GPS-class transceiver that has the ability to verify the position and contents of the SNF cask. The power conversion process, which converts the gamma photon energy into electrical power, is based on a variation of the successful dye-sensitized solar cell (DSSC) design developed by Konarka Technologies, Inc. (KTI). In particular, the focus of the current research is to make direct use of the high-energy gamma photons emitted from SNF, coupled with a scintillator material to convert some of the incident gamma photons into photons having wavelengths within the visible region of the electromagnetic spectrum. The high-energy gammas from the SNF will generate some power directly via Compton scattering and the photoelectric effect, and the generated visible photons output from the scintillator material can also be converted to electrical power in a manner similar to that of a standard solar cell. Upon successful implementation of an energy conversion device based on this new gammavoltaic principle, this inherent power source could then be utilized within SNF storage casks to drive a tamper-proof, low-power, electronic detection/security monitoring system for the spent fuel. The current project has addressed several aspects associated with this new energy conversion concept, including the development of a base conceptual design for an inherent gamma-induced power conversion unit for SNF monitoring, the characterization of the radiation environment that can be expected within a typical SNF storage system, the initial evaluation of Konarka's base solar cell design, the design and

  15. Plant Design for the Production of DUAGG

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.

    2003-02-04

    The cost of producing DUAGG is an important consideration for any interested private firm in determining whether DUCRETE is economically viable as a material of construction in next-generation spent nuclear fuel casks. This study analyzed this project as if it was a stand-alone project. The capital cost includes engineering design, equipment costs and installation, start up, and management; the study is not intended to be a life-cycle cost analysis. The costs estimated by this study are shown in Table ES.1, and the conclusions of this study are listed in Table ES.2. The development of DUAGG and DUCRETE is a major thrust of the Depleted Uranium Uses Research and Development Project. An obvious use of depleted uranium is as a shielding material (e.g., DUCRETE). DUCRETE is made by replacing the conventional stone aggregate in concrete with DUAGG. One objective of this project is to bring the development of DUCRETE to a point at which a demonstrated basis exists for its commercial deployment. The estimation of the costs to manufacture DUAGG is an important part of this effort. Paul Lessing and William Quapp developed DUAGG and DUCRETE as part of an Idaho National Engineering and Environmental Laboratory (INEEL) program to find beneficial uses for depleted uranium (DU). Subsequently, this technology was licensed to Teton Technologies, Inc. The DUAGG process mixes DUO{sub 2} with sintering materials and additives to form pressed briquettes. These briquettes are sintered at 1300 C, and the very dense sintered briquettes are then crushed and classified into gap-graded size fractions. The graded DUAGG is then ready to be used to make high-strength heavy DUCRETE. The DUCRETE shielding will be placed into an annular steel cask-shell mold, which has internal steel reinforcing bars. The objectives of this study are to (1) use previous DUAGG process developments to design a plant that will produce DUAGG at a baseline rate, (2) determine the size of the equipment required to meet

  16. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the &apos

  17. Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

    1999-12-01

    Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models

  18. Draft Geologic Disposal Requirements Basis for STAD Specification

    Energy Technology Data Exchange (ETDEWEB)

    Ilgen, Anastasia G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-25

    This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

  19. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  20. Alcohol in moderation

    DEFF Research Database (Denmark)

    Mueller, Simone; Lockshin, Larry; Louviere, Jordan J.

    2011-01-01

    Purpose: The study examines the market potential for low and very low alcohol wine products under two different tax regimes. The penetration and market share of low alcohol wine are estimated under both tax conditions. Consumers’ alcoholic beverage purchase portfolios are analysed and those...... products identified, which are jointly purchased with low alcohol wines. The effect of a tax increase on substitution patterns between alcoholic beverages is examined. Methodology: In a discrete choice experiment, based on their last purchase, consumers select one or several different alcoholic beverages...... volume is estimated under the current tax regime. Between six to eight percent of consumers are expected to adopt low alcohol wine alternatives as part of their alcoholic beverage portfolio. Consumers of cask wine and light beer are more likely and consumers of medium-full strength beer and spirits...

  1. Industrial research for transmutation scenarios

    Science.gov (United States)

    Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

    2011-04-01

    This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

  2. Radioactive waste material melter apparatus

    Science.gov (United States)

    Newman, D.F.; Ross, W.A.

    1990-04-24

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  3. Conceptual design of the handling and storage system for spent target vessel

    Energy Technology Data Exchange (ETDEWEB)

    Adachi, Junichi; Sasaki, Shinobu; Kaminaga, Masanori; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    A conceptual design of a handling and storage system for spent target vessels has been carried out, in order to establish spent target technology for the neutron scattering facility. The spent target vessels must be treated remotely with high reliability and safety, since they are highly activated and contain the poisonous mercury. The system is composed of a target exchange trolley to exchange the target vessel, remote handling equipment such as manipulators, airtight casks for the spent target vessel, storage pits and so on. This report presents the results of conceptual design study on a basic plan, a handling procedure, main devices and their arrangement of a handling and storage system for the spent target vessels. (author)

  4. Siegel[JMMM 7,312(`78)] FIRST EXPERIMENTAL DISCOVERY of Giant-Magnetoresistance Decade Pre ``Fert'' and ``Gruenberg'' ['88 - `78] = 10-Years = One-Decade Sounds, for Nuclear-Power Naïve ``Panacea'' for Global-Warming/Climate-Chan

    Science.gov (United States)

    Hoffmann, Masterace; Siegel, Edward

    Siegel[JMMM 7,312(`78); Monju (12/'95) LMFBR PREDICTION!!!] following: Wigner[JAP 17,857(`46)]-(Alvin)Weinberg(ANL/ORNL/ANS)-(Sidney)Siegel(ANL/ORNL/ANS)-Seitz-Overhauser-Rollnick-Pollard-Lofaro-Markey-Pringle[Nuclear-PowerFrom Physics to Politics(`79)] FIRST EXPERIMENTAL DISCOVERY [SiegelTHERMAL-leading-to-mechanical (TLTM)-INstability/``sensitization'' in: nuclear-reactors/spent-fuel dry-casks/refineries/jet/missile/rocket-engines/...SOUNDS A DIRE WARNING FOR NAIVE Hansen-Sommerville-Holdren-DOE-NRC-OSTP-WNA-NEI-AIP-APS-...calls/media-hype/P.R./spin-doctoring for carbon-``free'' nuclear-power as a SUPPOSED ``panacea'' for climate-change/global-warming: ``TRUST BUT VERIFY!!!'' ; a VERY LOUD CAVEAT EMPTOR!!!

  5. Radiation protection design of the Super-Fragment-Separator at the FAIR facility; Strahlenschutzplanung fuer den Super-Fragment-Separator bei FAIR

    Energy Technology Data Exchange (ETDEWEB)

    Lang, R.; Conrad, I.; Kozlova, E.; Radon, T.; Stuman, F.; Weick, H.; Winkler, M. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Fehrenbacher, G. [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); FAIR - Facility for Antiproton and Ion Research in Europe GmbH, Darmstadt (Germany)

    2015-07-01

    The Super FRagment- Separator (Super-FRS) at FAIR will produce atomic nuclei with an exotic ratio of protons to neutrons and demands highest standards of the constructional shielding and the handling process of activated components due to the required high intensities of the ion beam. The relevant sources of radiation, especially neutron radiation, resulting from the interaction of the accelerated ion beam with matter will be explained. In the vicinity of the pre-separator beam line, a high level of radioactive activation of the target (approx. 10{sup 11} Bq) and the beamcatcher (approx. 10{sup 13} Bq) is expected. Furthermore, the handling of activated components is explained and the concept of component handling and service with a specially manufactured shielding cask is also an important part of the ongoing design process. Finally, the hot cell, which is located in the same building, is presented. Radioactive components will be serviced, stored or conditioned for disposal therein.

  6. Status of ITER neutral beam cell remote handling system

    CERN Document Server

    Sykes, N; Choi, C-H; Crofts, O; Crowe, R; Damiani, C; Delavalle, S; Meredith, L; Mindham, T; Raimbach, J; Tesini, A; Van Uffelen, M

    2013-01-01

    The ITER neutral beam cell will contain up to three heating neutral beams and one diagnostic neutral beam, and four upper ports. Though manual maintenance work is envisaged within the cell, when containment is breached, or the radiological protection is removed the maintenance must be conducted remotely. This maintenance constitutes the removal and replacement of line replaceable units, and their transport to and from a cask docked to the cell. A design of the remote handling system has been prepared to concept level which this paper describes including the development of a beam line transporter, beam source remote handling equipment, upper port remote handling equipment and equipment for the maintenance of the neutral shield. This equipment has been developed complete the planned maintenance tasks for the components of the neutral beam cell and to have inherent flexibility to enable as yet unforeseen tasks and recovery operations to be performed.

  7. [Body thermal status under low-temperature conditions in brewing production].

    Science.gov (United States)

    Vasileva-Todorova, L; Dimitrova-Toneva, I

    1992-01-01

    The purpose of the present study is to trace the thermal state of workers exposed to low temperatures in brewery production, establishing the heat loss and the stress of thermoregulation. The investigations are performed in the departments for fermentation, deposit, cask washing and filling of 3 brewery plants. In order to characterize the microclimate methods of thermometry, psychometry and catathermometry are used. The heat state is controlled by methods of subjective heat perception, skin temperature, average skin temperature, temperature gradients, oral, rectal and average body temperature and the thermal content. The results of the physiological examinations point out to significant loss, which affects not only the periphery but also the deep tissues. There is an expressed risk of supercooling of the organism. The data of the heat deficit impose a correction of the working clothes and limitation of the exposure.

  8. The Microbial Diversity of Traditional Spontaneously Fermented Lambic Beer

    Science.gov (United States)

    Spitaels, Freek; Wieme, Anneleen D.; Janssens, Maarten; Aerts, Maarten; Daniel, Heide-Marie; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

    2014-01-01

    Lambic sour beers are the products of a spontaneous fermentation that lasts for one to three years before bottling. The present study determined the microbiota involved in the fermentation of lambic beers by sampling two fermentation batches during two years in the most traditional lambic brewery of Belgium, using culture-dependent and culture-independent methods. From 14 samples per fermentation, over 2000 bacterial and yeast isolates were obtained and identified. Although minor variations in the microbiota between casks and batches and a considerable species diversity were found, a characteristic microbial succession was identified. This succession started with a dominance of Enterobacteriaceae in the first month, which were replaced at 2 months by Pediococcus damnosus and Saccharomyces spp., the latter being replaced by Dekkera bruxellensis at 6 months fermentation duration. PMID:24748344

  9. Los Alamos Hot-Cell-Facility modifications for examining FFTF fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, B.M.; Ledbetter, J.M.

    1982-01-01

    Commissioned in 1960, the Wing 9 Hot Cell Facility at Los Alamos was recently modified to meet the needs of the 1980s. Because fuel pins from the Fast Flux Test Facility (FFTF) at the Hanford Engineering Development Laboratory (HEDL) are too long for examination in the original hot cells, we modified cells to accommodate longer fuel pins and to provide other capabilities as well. For instance, the T-3 shipping cask now can be opened in an inert atmosphere that can be maintained for all nondestructive and destructive examinations of the fuel pins. The full-length pins are visually examined and photographed, the wire wrap is removed, and fission gas is sampled. After the fuel pin is cropped, a cap is seal-welded on the section containing the fuel column. This section is then transferred to other cells for gamma-scanning, radiography, profilometry, sectioning for metallography, and chemical analysis.

  10. Criticality safety evaluation report for the multi-canister overpack

    Energy Technology Data Exchange (ETDEWEB)

    KESSLER, S.F.

    1999-05-21

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark 1V or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operations at the Cold Vacuum Drying Facility, and storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} = 0.95 criticality safety limit. Additional analyses in this revision include partial fuel basket loadings, loading 26.1 inch Mark IA fuel assemblies into Mark IV fuel baskets, and the revised fuel and scrap basket designs. The MCO MCNP model was revised to include the shield plug assembly.

  11. Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

    Energy Technology Data Exchange (ETDEWEB)

    KESSLER, S.F.

    1999-10-20

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weld station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.

  12. Evolution of the MACSTOR{trademark} dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Joubert, W.M. [Atomic Energy of Canada Ltd., Montreal, Quebec (Canada)

    1995-12-31

    The MACSTOR{trademark} (Modular Air-Cooled Canister Storage) system was developed by Atomic Energy of Canada Limited (AECL) for the interim storage of spent fuel discharged by light water reactors. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The system includes all the necessary equipment to transfer spent fuel from a storage pool to an independent interim dry spent fuel storage site. After presenting a description of the system and a brief history of its development, the paper addresses its thermal performance and modeling for various design configurations. Finally, a brief summary of the experience being gained during the implementation of a MACSTOR{trademark} system modified for CANDU spent fuel at the Gentilly-2 NPP in Quebec is presented. It includes progress made in licensing activities and in public hearings pertinent to the initiation of the project.

  13. Analysis of radwaste material management options for experimental DUPIC fuel fabrication process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Yang, M. S.; Kim, K. H.; Shin, J. M.; Lee, H. S.; Ko, W. I.; Lee, J. W.; Yim, S. P.; Hong, D. H.; Lee, J. Y.; Baik, S. Y.; Song, W. S.; Yoo, B. O.; Lee, E. P.; Kang, I. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This report is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This report was written for helping researchers working in related facilities to DUPIC project understanding management of DUPIC radioactive waste as well as fellows in DUPIC project. Also, it will be used as basic material to prove transparency and safeguardability of DUPIC fuel cycle. In order to meet these purposes, this report includes basic experiment plan for manufacturing DUPIC nuclear fuel, outlines for DUPIC manufacturing facility and equipment, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures. 15 refs., 31 figs., 11 tabs. (Author)

  14. Thermal evaluation facility for LMFBR spent fuel transport

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations.

  15. Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

  16. Radiation safety in sea transport of radioactive material in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Odano, N. [National Maritime Research Inst., Tokyo (Japan); Yanagi, H. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan)

    2004-07-01

    Radiation safety for sea transport of radioactive material in Japan has been discussed based on records of the exposed dose of sea transport workers and measured data of dose rate equivalents distribution inboard exclusive radioactive material shipping vessels. Recent surveyed records of the exposed doses of workers who engaged in sea transport operation indicate that exposed doses of transport workers are significantly low. Measured distribution of the exposed dose equivalents inboard those vessels indicates that dose rate equivalents inside those vessels are lower than levels regulated by the transport regulations of Japan. These facts clarify that radiation safety of inboard environment and handling of transport casks in sea transport of radioactive material in Japan are assured.

  17. Transportation Shock and Vibration Literature Review

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Lahti, Erik A.; Ross, Steven B.

    2013-06-06

    This report fulfills the M4 milestone M4FT-13OR08220112, "Report Documenting Experimental Activities." The purpose of this report is to document the results of a literature review conducted of studies related to the vibration and shock associated with the normal conditions of transport for rail shipments of used nuclear fuel from commercial light-water reactors. As discussed in Adkins (2013), the objective of this report is to determine if adequate data exist that would enable the impacts of the shock and vibration associated with the normal conditions of transport on commercial light-water reactor used nuclear fuel shipped in current generation rail transportation casks to be realistically modeled.

  18. Siegel FIRST EXPERIMENTAL DISCOVERY of Granular-Giant-Magnetoresistance (G-GMR) DiagnosES/ED Wigner's-Disease/.../Spinodal-Decomposition in ``Super''Alloys Generic Endemic Extant in: Nuclear-Reactors/ Petrochemical-Plants/Jet/ Missile-Engines/...

    Science.gov (United States)

    Hoffman, Ace; Wigner-Weinberg, Eugene-Alvin; Siegel, Edward Carl-Ludwig Sidney; ORNL/Wigner/Weinberg/Siegel/Hollifeld/Yu/... Collaboration; ANL/Fermi/Wigner/Arrott/Weeks/Bader/Freeman/Sinha/Palazlotti/Nichols/Petersen/Rosner/Zimmer/... Collaboration; BNL/Chudahri/Damask/Dienes/Emery/Goldberg/Bak//Bari/Lofaro/... Collaboration; LLNL-LANL/Hecker/Tatro/Meara/Isbell/Wilkins/YFreund/Yudof/Dynes/Yang/... Collaboration; WestinKLouse/EPRI/PSEG/IAEA/ABB/Rickover/Nine/Carter/Starr/Stern/Hamilton/Richards/Lawes/OGrady/Izzo Collaboration

    2013-03-01

    Siegel[APS Shock-Physics Mtg., Chicago(11)] carbides solid-state chemistry[PSS (a)11,45(72); Semis. & Insuls. 5: 39,47,62 (79)], following: Weinberg-Siegel-Loretto-Hargraves-Savage-Westwood-Seitz-Overhauser-..., FIRST EXPERIMENTAL DISCOVERY of G-GMR[JMMM 7, 312(78); Google: ``If LEAKS Could KILL Ana Mayo''] identifIED/IES GENERIC ENDEMIC EXTANT domination of old/new (so mis-called) ``super''alloys': nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe/Co-based (so mis-called) ''super''alloys (182/82; Hastelloy-X,600,304/304L-Stainless-Steels,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms!!!): THERMAL: Wigner's-disease(WD physics) [J.Appl.Phys.17,857(46)]/ Ostwald-ripening

  19. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  20. Evaluation of the active odorants in Amontillado sherry wines during the aging process.

    Science.gov (United States)

    Moyano, Lourdes; Zea, Luis; Moreno, Jose A; Medina, Manuel

    2010-06-09

    Odor compounds in Amontillado sherry white wine obtained by means of biological aging first and oxidative aging second in American oak casks were determined by gas chromatography-olfactometry. Sniffing revealed fruity, fatty, chemical, spicy, vegetable, floral and empyreumatic odors, the first being the most common. Olfactometric intensity was assessed on a four-point scale. Most changes were detected during the first years of the oxidative aging step. Ethyl isobutanoate, ethyl butanoate, ethyl octanoate, and eugenol were the strongest odor compounds detected by sniffing in wines. The odor spectrum values for all active odorants were calculated in relation to ethyl octanoate, this compound being the most potent odorant. On the basis of olfactometric intensities and odor spectrum values, ethyl octanoate, ethyl butanoate, eugenol, ethyl isobutanoate, and sotolon can be deemed the main group of potent odorants in Amontillado wines. These compounds maintained similar relative contributions to the aroma profile during the oxidative aging step.

  1. Synthesis of uranium metal using laser-initiated reduction of uranium tetrafluoride by calcium metal

    Energy Technology Data Exchange (ETDEWEB)

    West, M.H.; Martinez, M.M.; Nielsen, J.B.; Court, D.C.; Appert, Q.D.

    1995-09-01

    Uranium metal has numerous uses in conventional weapons (armor penetrators) and nuclear weapons. It also has application to nuclear reactor designs utilizing metallic fuels--for example, the former Integral Fast Reactor program at Argonne National Laboratory. Uranium metal also has promise as a material of construction for spent-nuclear-fuel storage casks. A new avenue for the production of uranium metal is presented that offers several advantages over existing technology. A carbon dioxide (CO{sub 2}) laser is used to initiate the reaction between uranium tetrafluoride (UF{sub 4}) and calcium metal. The new method does not require induction heating of a closed system (a pressure vessel) nor does it utilize iodine (I{sub 2}) as a chemical booster. The results of five reductions of UF{sub 4}, spanning 100 to 200 g of uranium, are evaluated, and suggestions are made for future work in this area.

  2. Norauto. Environment report 2000; Norauto. Environnement rapport 2000

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    Norauto is the French leader of auto repair centers. It is today an international company dealing with car maintenance and parts and established in seven countries. For several years, the company has participated in the fight against pollution linked to motor vehicle activity with the contribution to the sorting, collection and re-use of wastes relating to vehicles. This environmental report takes stock of Norauto's recycling activity in the domain of batteries, tires, lubrication oils, oil and fuel filters, packaging (casks and cans), cooling liquid, and small and telephone batteries and electronic wastes. New waste management channels have been established in 2000 and concern the selective sorting of paper, the recycling of rags and the maintenance of storm sewer systems of Norauto centers. Specific fountains have been designed for the cleaning of breaking systems and to avoid the release of aerosol products to the atmosphere. (J.S.)

  3. Performance of the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities.

    Science.gov (United States)

    Lahaye, T; Chau, Q; Ménard, S; Lacoste, V; Muller, H; Luszik-Bhadra, M; Reginatto, M; Bruguier, P

    2006-01-01

    This paper mainly aims at presenting the measurements and the results obtained with the electronic personal neutron dosemeter Saphydose-N at different facilities. Three campaigns were led in the frame of the European contract EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The first one consisted in the measurements at the IRSN French research laboratory in reference neutron fields generated by a thermal facility (SIGMA), radionuclide ISO sources ((241)AmBe; (252)Cf; (252)Cf(D(2)O)\\Cd) and a realistic spectrum (CANEL/T400). The second one was performed at the Krümmel Nuclear Power Plant (Germany) close to the boiling water reactor and to a spent fuel transport cask. The third one was realised at Mol (Belgium), at the VENUS Research Reactor and at Belgonucléaire, a fuel processing factory.

  4. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  5. Information handbook on independent spent fuel storage installations

    Energy Technology Data Exchange (ETDEWEB)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  6. HAZARDS OF THERMAL EXPANSION FOR RADIOLOGICAL CONTAINER ENGULFED IN FIRE

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2013-05-01

    Fire accidents pose a serious threat to nuclear facilities. It is imperative that transport casks or shielded containers designed to transport/contain radiological materials have the ability to withstand a hypothetical fire. A numerical simulation was performed for a shielded container constructed of stainless steel and lead engulfed in a hypothetical fire as outlined by 10 CFR §71.73. The purpose of this analysis was to determine the thermal response of the container during and after the fire. The thermal model shows that after 30 minutes of fire, the stainless steel will maintain its integrity and not melt. However, the lead shielding will melt since its temperature exceeds the melting point. Due to the method of construction of the container under consideration, ample void space must be provided to allow for thermal expansion of the lead upon heating and melting, so as to not overstress the weldment.

  7. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  8. Rail Access to Yucca Mountain: Critical Issues

    Energy Technology Data Exchange (ETDEWEB)

    Halstead, R. J.; Dilger, F.; Moore, R. C.

    2003-02-25

    The proposed Yucca Mountain repository site currently lacks rail access. The nearest mainline railroad is almost 100 miles away. Absence of rail access could result in many thousands of truck shipments of spent nuclear fuel and high-level radioactive waste. Direct rail access to the repository could significantly reduce the number of truck shipments and total shipments. The U.S. Department of Energy (DOE) identified five potential rail access corridors, ranging in length from 98 miles to 323 miles, in the Final Environmental Impact Statement (FEIS) for Yucca Mountain. The FEIS also considers an alternative to rail spur construction, heavy-haul truck (HHT) delivery of rail casks from one of three potential intermodal transfer stations. The authors examine the feasibility and cost of the five rail corridors, and DOE's alternative proposal for HHT transport. The authors also address the potential for rail shipments through the Las Vegas metropolitan area.

  9. A unique radioisotopic label as a new concept for safeguarding and tagging of long-term stored items and waste

    CERN Document Server

    Chernikova, Dina

    2013-01-01

    The present paper discuss a novel method of tagging and labeling of waste casks, copper canisters, spent fuel containers, mercury containers, waste pack- ages and other items. In particular, it is related to the development of new long-term security identification tags/labels that can be applied to articles for carrying information about the content, inventory tracking, prevention of falsifi- cation and theft etc. It is suggested to use a unique combination of radioisotopes with different predictable length of life, as a label of the items. The possibil- ity to realize a multidimensional bar code symbology is proposed as an option for a new labeling method. The results of the first tests and evaluations of this are shown and discussed in the paper. The invention is suitable for use in items assigned to long-term (hundreds of years) storing or for final repositories. Alternative field of use includes fresh nuclear fuel handling and shipment of goods.

  10. Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, P.F.

    1994-10-13

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment.

  11. Monte Carlo importance sampling for the MCNP{trademark} general source

    Energy Technology Data Exchange (ETDEWEB)

    Lichtenstein, H.

    1996-01-09

    Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence.

  12. Draft fracture mechanics code case for American Society of Mechanical Engineers NUPACK rules

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, P.; Sorenson, K. [Sandia National Labs., Albuquerque (United States); Nickell, R. [Applied Science and Technology, Poway (United States); Saegusa, T. [Central Research Inst. for Electric Power Industry, Abiko (Japan)

    2004-07-01

    The containment boundaries of most spent-fuel casks certified for use in the United States by the Nuclear Regulatory Commission are constructed with stainless steel, a material that is ductile in an engineering sense at all temperatures and for which, therefore, fracture mechanics principles are not relevant for the containment application. Ferritic materials may fail in a nonductile manner at sufficiently low temperatures, so fracture mechanics principles may be applied to preclude nonductile fracture. Because of the need to transport and store spent nuclear fuel safely in all types of climatic conditions, these vessels have regulatory lowest service temperatures that range down to -40 C (-40 F) for transport application. Such low service temperatures represent a severe challenge in terms of fracture toughness to many ferritic materials. Linear-elastic and elastic-plastic fracture mechanics principles provide a methodology for evaluating ferritic materials under such conditions.

  13. Effect of an excise tax increase on the demand for low alcohol wine

    DEFF Research Database (Denmark)

    Mueller, Simone; Lockshin, Larry; Louviere, Jordan J.

    alcohol strength is seen as particularly promising direction to enhance consumer choice for lower alcohol alternatives. Over the past twenty years, average alcohol strength of wine has increased due to a growing viticultural and oenological emphasis on riper fruit and more expressive taste profiles......Purpose: The study examines the demand for low and very low alcohol wine products under the current Australian volumetric alcohol system and under an increase of excise tax for wine under an alcohol tax equalisation. The penetration and market share of normal strength and low alcohol wine....... Background: Contrary to volumetric taxed spirits and beer, wine is currently taxed on a value basis in Australia. This has led to cask wine being the cheapest way to get drunk, resulting in high social and health problems, particularly in Australia’s aboriginal communities in the Northern Territories, where...

  14. MRS system study for the repository: Yucca Mountain Site Characterization Project; Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Sinagra, T.A. [Bechtel National, Inc., San Francisco, CA (USA); Harig, R. [Parsons, Brinckerhoff, Quade and Douglas, Inc., San Francisco, CA (USA)

    1990-12-01

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), has initiated a waste management system study to identify the impacts of the presence or absence of a monitored retrievable storage facility (hereinafter referred to as ``MRS``) on system costs and program schedules. To support this study, life-cycle cost estimates and construction schedules have been prepared for the surface and underground facilities and operations geologic nuclear waste repository at Yucca Mountain, Nye County, Nevada. Nine different operating scenarios (cases) have been identified by OCRWM for inclusion in this study. For each case, the following items are determined: the repository design and construction costs, operating costs, closure and decommissioning costs, required staffing, construction schedules, uncertainties associated with the costs and schedules, and shipping cask and disposal container throughputs. This document contains A-D.

  15. Synchronous two-dimensional MIR correlation spectroscopy (2D-COS) as a novel method for screening smoke tainted wine.

    Science.gov (United States)

    Fudge, Anthea L; Wilkinson, Kerry L; Ristic, Renata; Cozzolino, Daniel

    2013-08-15

    In this study, two-dimensional correlation spectroscopy (2D-COS) combined with mid-infrared (MIR) spectroscopy was evaluated as a novel technique for the identification of spectral regions associated with smoke-affected wine, for the purpose of screening taint arising from grapevine exposure to smoke. Smoke-affected wines obtained from experimental and industry sources were analysed using MIR spectroscopy and chemometrics, and calibration models developed. 2D-COS analysis was used to generate synchronous data maps for red and white cask wines spiked with guaiacol, a marker of smoke taint. Correlations were observed at wavelengths that could be attributable to aromatic C-C stretching, i.e., between 1400 and 1500 cm(-1), indicative of volatile phenols. These results demonstrate the potential of 2D-COS as a rapid, high-throughput technique for the preliminary screening of smoke tainted wine.

  16. Status of ITER neutral beam cell remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Sykes, N., E-mail: nick.sykes@ccfe.ac.uk [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Belcher, C. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Choi, C.-H. [ITER Organisation, CS90 046, 13067 St. Paul les Durance Cedex (France); Crofts, O. [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Crowe, R. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Damiani, C. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Delavalle, S.; Meredith, L. [Oxford Technologies Ltd, Abingdon OX14 1RJ (United Kingdom); Mindham, T.; Raimbach, J. [CCFE. Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Tesini, A. [ITER Organisation, CS90 046, 13067 St. Paul les Durance Cedex (France); Van Uffelen, M. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain)

    2013-10-15

    The ITER neutral beam cell will contain up to three heating neutral beams and one diagnostic neutral beam, and four upper ports. Though manual maintenance work is envisaged within the cell, when containment is breached, or the radiological protection is removed the maintenance must be conducted remotely. This maintenance constitutes the removal and replacement of line replaceable units, and their transport to and from a cask docked to the cell. A design of the remote handling system has been prepared to concept level which this paper describes including the development of a beam line transporter, beam source remote handling equipment, upper port remote handling equipment and equipment for the maintenance of the neutral shield. This equipment has been developed complete the planned maintenance tasks for the components of the neutral beam cell and to have inherent flexibility to enable as yet unforeseen tasks and recovery operations to be performed.

  17. MRS system study for the repository: Yucca Mountain Site Characterization Project; Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Sinagra, T.A. [Bechtel National, Inc., San Francisco, CA (USA); Harig, R. [Parsons, Brinckerhoff, Quade and Douglas, Inc., San Francisco, CA (USA)

    1990-12-01

    The US Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), has initiated a waste management system study to identify the impacts of the presence or absence of a monitored retrievable storage facility (hereinafter referred to as ``MRS``) on system costs and program schedules. To support this study, life-cycle cost estimates and construction schedules have been prepared for the surface and underground facilities and operations of a geologic nuclear waste repository at Yucca Mountain, Nye County, Nevada. Nine different operating scenarios (cases) have been identified by OCRWM for inclusion in this study. For each case, the following items are determined: the repository design and construction costs, operating costs, closure and decommissioning costs, required staffing, construction schedules, uncertainties associated with the costs and schedules, and shipping cask and disposal container throughputs. 6 refs., 83 figs., 57 tabs.

  18. Concept design on RH maintenance of CFETR Tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Wu, Songtao; Wan, Yuanxi; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Ye, Minyou [University of Science and Technology of China, Hefei (China); Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-10-15

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

  19. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  20. Demonstrating the Safety of Long-Term Dry Storage - 13468

    Energy Technology Data Exchange (ETDEWEB)

    McCullum, Rod [Nuclear Energy Institute, 1201 F St. NW, Washington, DC, 20004 (United States); Brookmire, Tom [Dominion Energy, 5000 Dominion Boulevard Glen Allen, VA 23060 (United States); Kessler, John [Electric Power Research Institute, 1300 West W.T. Harris Boulevard, Charlotte, NC 28262 (United States); Leblang, Suzanne [Entergy, 1340 Echelon Parkway, Jackson, MS 39211 (United States); Levin, Adam [Exelon, 4300 Winfield Road, Warrenville, IL 60555 (United States); Martin, Zita [Tennessee Valley Authority, 1101 Market Street, Chattanooga, TN 37402 (United States); Nesbit, Steve [Duke Energy, 550 South Tryon Street, Charlotte, NC 28202 (United States); Nichol, Marc [Nuclear Energy Institute, 1201 F St. NW Washington DC, 2004 (United States); Pickens, Terry [Xcel Energy, 414 Nicollet Mall, Minneapolis, MN 55401 (United States)

    2013-07-01

    Commercial nuclear plants in the United States were originally designed with the expectation that used nuclear fuel would be moved directly from the reactor pools and transported off site for either reprocessing or direct geologic disposal. However, Federal programs intended to meet this expectation were never able to develop the capability to remove used fuel from reactor sites - and these programs remain stalled to this day. Therefore, in the 1980's, with reactor pools reaching capacity limits, industry began developing dry cask storage technology to provide for additional on-site storage. Use of this technology has expanded significantly since then, and has today become a standard part of plant operations at most US nuclear sites. As this expansion was underway, Federal programs remained stalled, and it became evident that dry cask systems would be in use longer than originally envisioned. In response to this challenge, a strong technical basis supporting the long term dry storage safety has been developed. However, this is not a static situation. The technical basis must be able to address future challenges. Industry is responding to one such challenge - the increasing prevalence of high burnup (HBU) used fuel and the need to provide long term storage assurance for these fuels equivalent to that which has existed for lower burnup fuels over the past 25 years. This response includes a confirmatory demonstration program designed to address the aging characteristics of HBU fuel and set a precedent for a learning approach to aging management that will have broad applicability across the used fuel storage landscape. (authors)

  1. Supply and demand for alcohol in Australia: relationships between industry structures, regulation and the marketplace.

    Science.gov (United States)

    Stockwell, T; Crosbie, D

    2001-07-01

    Aspects of alcohol supply and demand relationships are examined in relation to the two main beverage varieties in Australia, wine and beer. It is argued that this case study illustrates how the 'supply side' is able to create and protect demand for alcohol through both taking advantage of and influencing government regulation of the market for alcohol. In relation to low alcohol beer the impact on public health and safety has been extremely positive. In relation to the creation of cask wine in the late 1960s there have been demonstrably deleterious effects. Preferential taxation arrangements for the Australian wine industry have dramatically increased both exports and home consumption. One unintended consequence has been the creation of a major new market for cheap bulk wines that have had a devastating public health impact, particularly on Aboriginal communities, and also the invention of 'alco-pops'. Two-thirds of all table wine consumed in Australia now comes in a cask and 90% of this product is manufactured by three multi-national companies that wield enormous power and political influence to maintain the status quo. The Australian beer industry is well known internationally for its export of 'full strength' (around 5% by volume) beers. What is less well known is its commercial success in the development of low and mid-strength varieties for home consumption. In some States these now comprise 40% by volume of the beer market. This development can largely be attributed to State taxation arrangements, to drink-driving law enforcement, marketing strategies and to a decade of intense competition between several major brewers. This case study indicates how alcohol taxation policy can have major impacts on public health (both positive and negative) but that in a modern market economy it is difficult for governments to act in the public interest due to pressures from vested interest groups.

  2. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  3. Paradigm shift in transport legislation or rather at the 'bottleneck'; Paradigmenwechsel im Befoerderungsrecht oder am 'Flaschenhals'

    Energy Technology Data Exchange (ETDEWEB)

    Naeser, Hanns [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2015-01-15

    In the year just started significant decisions with considerable consequences by the Federal Constitutional Court and the Federal Administrative Court in the field of nuclear law are expected. Especially the decision with regards to 'nuclear phase-out' within the 13th amendment of the Atomic Energy Act is being eagerly expected, as with its far-reaching consequences also fundamental constitutional questions need to be answered. The Federal Administrative Court will need to decide on the question, whether she admits the appeal against the Brunsbuettel decision by the Higher Administrative Court Schleswig-Holstein (HAC), which from the view of claimant shifted the fundamental basis of demarcation of responsibilities between the executive and judiciary power. In comparison to these fundamental decisions the awaited decision by the HAC on nuclear transport legislation seems of subordinate importance, although she will proceed with a paradigm shift in the legal area. The decision deals with the question as to whether and when a right of action from a third party within the nuclear transport legislation can be accepted or more precisely under which preconditions a third party has clear standing against a nuclear transport authorisation. As the site selection law (issued on 23 July 2013 BGBI I p. 2552) excludes the recirculation of vitrified waste block canisters from reprocessing spent fuel elements to the transport cask storage facility Gorleben, the decision by the HAC Lueneburg for this site will only be relevant for present unpredictable transportations from the transport cask storage facility Gorleben to a final repository. If necessary interest to seek a declaratory judgment for declaratory action, in concreto danger of recurrence will be approved, is another matter.

  4. Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Wagner, John C [ORNL

    2013-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

  5. Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred C. [Black Mountain Research, Henderson, NV 81012 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)

    2013-07-01

    As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

  6. 不同生态修复措施对滇池沉积物无机氮迁移转化影响%Effects of Different Ecological Restoration Measures on Transferring and Transformation of Inorganic Nitrogen in Dianchi Sediment

    Institute of Scientific and Technical Information of China (English)

    陈国元; 周易勇

    2011-01-01

    Effects of three ecological restoration measures which are planting Vallisneria spiraslis L. directly in Dianchi sediment, planting Vallisneria spiraslis L. in sediment covered by silver sand and planting Vallisneria spiraslis L. in sediment covered by laterite on transfer and transformation of inorganic nitrogen were studied by indoor incubation experiment. Results showed that the concentrations of CODMn, SS and NH4+-N in waters and NH/-N in sediment were the highest in plastic casks treated with planting Vallisneria spiraslis L. diretly in sediment, and there conentrations in plastic casks treated with planting Vallisneria spiraslis L. in sediment covered by laterite were the lowest. Ammonia and nitrite oxidation in waters firstly appear in the plastic casks treated with planting Vallisneria spiraslis L. in sediment covered by silver sand and laterite, and the date were the 13th and 19th day separately. Rates of ammonia and nitrite oxidation in waters and sediment were the highest in plastic casks treated with planting Vallisneria spiraslis L. in sediment covered by laterite, and followed by rates in plastic casks treated with planting Vallisneria spiraslis L. in sediment covered by silver sand. Results indicated that planting Vallisneria spiraslis L. in Dianchi sediment covered by silver sand and laterite are able to not only hold back the release of reductive material, suspended paniculate material and ammonia into water from sediment, but also decrease markedly ammonia concentration in sediment. Planting Vallisneria spiraslis L. in sediment covered by laterite which provided an effective attachment matrix for Vallisneria spiraslis L. to take a root and promoted Vallisneria spiraslis L. grow and recovery has the better result than planting Vallisneria spiraslis L. in sediment covered by silver sand, which indicated that planting Vallisneria spiraslis L. in sediment covered by laterite is the effective way of ecological restoration for contaminative sediment.%

  7. Structural Sensitivity of Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Karri, Naveen K.; Adkins, Harold E.; Hanson, Brady D.

    2013-09-27

    This LS-DYNA modeling study evaluated a generic used nuclear fuel vertical dry storage cask system under tip-over, handling drop, and seismic load cases to determine the sensitivity of the canister containment boundary to these loads. The goal was to quantify the expected failure margins to gain insight into what material changes over the extended long-term storage lifetime could have the most influence on the security of the containment boundary. It was determined that the tip-over case offers a strong challenge to the containment boundary, and identifies one significant material knowledge gap, the behavior of welded stainless steel joints under high-strain-rate conditions. High strain rates are expected to increase the material’s effective yield strength and ultimate strength, and may decrease its ductility. Determining and accounting for this behavior could potentially reverse the model prediction of a containment boundary failure at the canister lid weld. It must be emphasized that this predicted containment failure is an artifact of the generic system modeled. Vendor specific designs analyze for cask tip-over and these analyses are reviewed and approved by the Nuclear Regulatory Commission. Another location of sensitivity of the containment boundary is the weld between the base plate and the canister shell. Peak stresses at this location predict plastic strains through the whole thickness of the welded material. This makes the base plate weld an important location for material study. This location is also susceptible to high strain rates, and accurately accounting for the material behavior under these conditions could have a significant effect on the predicted performance of the containment boundary. The handling drop case was largely benign to the containment boundary, with just localized plastic strains predicted on the outer surfaces of wall sections. It would take unusual changes in the handling drop scenario to harm the containment boundary, such as

  8. Regulatory oversight report 2012 concerning nuclear safety in Swiss nuclear installations; Aufsichtsbericht 2012 zur nuklearen Sicherheit in den schweizerischen Kernanlagen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-15

    The Swiss Federal Nuclear Safety Inspectorate (ENSI) assesses and monitors nuclear facilities in Switzerland. These include the five nuclear power plants, the interim storage facilities based at each plant, the Central Interim Storage Facility (ZWILAG) and the nuclear facilities at the Paul Scherrer Institute (PSI), at the Federal Institute of Technology in Lausanne (EPFL) and at the University of Basel. Using a combination of inspections, regulatory meetings, examinations and analyses together with reports from the licensees of individual facilities, ENSI obtains the required overview of nuclear safety in the relevant facilities. It ensures that the facilities comply with the regulations and operate as required by law. Its regulatory responsibilities also include the transport of radioactive materials from and to nuclear facilities and the preparations for a deep geological repository for nuclear waste. ENSI maintains its own emergency organisation. It formulates and updates its own guidelines which stipulate the criteria for evaluating the current activities and future plans of the operators of nuclear facilities. ENSI produces regular reports on its regulatory activities and nuclear safety in Swiss nuclear facilities. It fulfils its statutory obligation to provide the public with information on particular events and findings in nuclear facilities. In 2012, the five nuclear power plants in Switzerland were all operated safely. 34 events were reported; on the international INES scale of 0 to 7, ENSI rated 33 events as Level 0 and 1 as Level 1. ENSI evaluates the safety of each nuclear power plant as part of a systematic safety evaluation taking account of both reportable events and other findings, in particular the results of more than 400 inspections conducted by ENSI during 2012. ZWILAG consists of several interim storage halls, a conditioning plant and an incineration/melting plant. At the end of 2012, the cask storage hall contained 40 transport/storage casks

  9. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the

  10. Safety relevant aspects of the long-term intermediate storage of spent fuel elements and vitrified high-level radioactive wastes; Sicherheitstechnische Aspekte der langfristigen Zwischenlagerung von bestrahlten Brennelementen und verglastem HAW

    Energy Technology Data Exchange (ETDEWEB)

    Ellinger, A.; Geupel, S.; Gewehr, K.; Gmal, B.; Hannstein, V.; Hummelsheim, K.; Kilger, R.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schmidt, G.; Spieth-Achtnich, A. [Oeko-Institut e.V. - Institut fuer angewandte Oekolgie (Germany)

    2010-04-15

    The currently in Germany pursued concept for management of spent fuel from nuclear power plants provides intermediate dry cask storage at the NPP sites until direct disposal in a deep geologic repository. In addition the earlier commissioned centralized dry storage facilities are being used for storage of high level radioactive waste returned from foreign reprocessing of German spent fuel performed so far. The dry interim storage facilities are licensed for 40 years of operation time. According to the German regulations a full scope periodic safety review is not required so far, neither practical experience on dry storage for this period of time is available. With regard to this background the report at hand is dealing with long term effects, which may affect safety of the interim storage during the 40 years period or beyond if appropriate, and with the question, whether additional analyses or monitoring measures may be required. Therefore relevant publications have been evaluated, calculations have been performed as well as a systematic screening with regard to loads and possible ageing effects has been applied to structures and components important for safety of the storage, in order to identify relevant long term effects, which may not have been considered sufficiently so far and to provide proposals for an improved ageing management. The report firstly provides an overview on the current state of technology describing shortly the national and international practice and experience. In the following chapters safety aspects of interim storage with regard to time dependent effects and variations are being analyzed and discussed. Among this not only technical aspects like the long term behavior of fuel elements, canisters and storage systems are addressed, but also operational long term aspects regarding personnel planning, know how conservation, documentation and quality management are taken into account. A separate chapter is dedicated to developing and describing

  11. Fuel Aging in Storage and Transportation (FAST): Accelerated Characterization and Performance Assessment of the Used Nuclear Fuel Storage System

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States)

    2016-08-02

    corrosion degradation of the UNF canisters in various environments. The methods employed included: 1) electrochemical corrosion testing, 2) stress corrosion cracking (SCC) studies, 3) characterization of the corrosion of welds, bolted joints, and seals under prototypic conditions, and 4) integration the results of corrosion tests and microstructural analyses. Technical Mission Area 4: Novel System Monitoring The focus of this mission was to develop in-situ prognostic monitoring methods to verify the condition of the UNF and the internal canister environment without having to break the seal of the inner cask. This included the development of self-powered sensing methods to continuously monitor the cask system and provide physical, safety, and security information during

  12. Investigation of properties of (U)HMW-PE under the influence of gamma irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ehe, K. von der; Wolff, D.; Boehning, M. [BAM Federal Institute for Materials Research and Testing, Berlin (Germany)

    2012-11-01

    High molecular weight polyethylene (HMW-PE) and ultra high molecular weight polyethylene (UHMW-PE) are high performance materials, which are predestinated for a wide range of applications due to characteristics like high chemical resistance, low wear, and high hydrogen content. In addition to the application of (U)HMW-PE as sports equipment (sliding surface of snowboards, skis) and in mechanical engineering (materials for spur gears and chain guides) these polymers are used in two other and even more demanding areas: as implant material in medical technology and as a component for neutron shielding purposes in casks for storage and transport of radioactive waste. In these two cases, gamma irradiation and its impact on the molecular structure of polyethylene play an important role. In the first case, irradiation is applied for sterilization purposes and in the second one it exists as a side-effect of inserting the radioactive material in the cask. Specifically concerning their use in the field of medical technology, different types of UHMW-PE have been intensively studied. It is generally accepted that two concurrent and competitive processes, based on initial chain scission and subsequent reactions of C-centered radicals and molecular fragments, occur in PE as a consequence of radiation: radical recombination accounts for crosslinking, together with some disproportionation, formation of low molecular weight fragments, and recrystallization. Furthermore, formation of oxygenated structures in the presence of traces of oxygen is a competitive process to those mentioned above. Radiation induced scission preferentially takes place in the amorphous phase and noncrystalline surface of the crystallites and results in shorter polymer chains, fewer entanglements, and consequently, increased molecular mobility. The latter enables folding of polymer segments and thus, crystal growth and increased perfection in the crystal lamellae. With regard to the long-term application of

  13. SLUDGE TREATMENT PROJECT ENGINEERED CONTAINER RETRIEVAL AND TRANSFER SYSTEM PRELMINARY DESIGN HAZARD AND OPERABILITY STUDY

    Energy Technology Data Exchange (ETDEWEB)

    CARRO CA

    2011-07-15

    This Hazard and Operability (HAZOP) study addresses the Sludge Treatment Project (STP) Engineered Container Retrieval and Transfer System (ECRTS) preliminary design for retrieving sludge from underwater engineered containers located in the 105-K West (KW) Basin, transferring the sludge as a sludge-water slurry (hereafter referred to as 'slurry') to a Sludge Transport and Storage Container (STSC) located in a Modified KW Basin Annex, and preparing the STSC for transport to T Plant using the Sludge Transport System (STS). There are six, underwater engineered containers located in the KW Basin that, at the time of sludge retrieval, will contain an estimated volume of 5.2 m{sup 3} of KW Basin floor and pit sludge, 18.4 m{sup 3} of 105-K East (KE) Basin floor, pit, and canister sludge, and 3.5 m{sup 3} of settler tank sludge. The KE and KW Basin sludge consists of fuel corrosion products (including metallic uranium, and fission and activation products), small fuel fragments, iron and aluminum oxide, sand, dirt, operational debris, and biological debris. The settler tank sludge consists of sludge generated by the washing of KE and KW Basin fuel in the Primary Clean Machine. A detailed description of the origin of sludge and its chemical and physical characteristics can be found in HNF-41051, Preliminary STP Container and Settler Sludge Process System Description and Material Balance. In summary, the ECRTS retrieves sludge from the engineered containers and hydraulically transfers it as a slurry into an STSC positioned within a trailer-mounted STS cask located in a Modified KW Basin Annex. The slurry is allowed to settle within the STSC to concentrate the solids and clarify the supernate. After a prescribed settling period the supernate is decanted. The decanted supernate is filtered through a sand filter and returned to the basin. Subsequent batches of slurry are added to the STSC, settled, and excess supernate removed until the prescribed quantity of sludge is

  14. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Loiseau, O. (Institut de radioprotection et de Surete Nucleaire, France); Koch, W. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Billone, M. C. (Argonne National Laboratory, USA); Lucero, Daniel A.; Burtseva, T. (Argonne National Laboratory, USA); Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO{sub 2}, test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence

  15. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design.

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has

  16. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  17. An approach to determine a defensible spent fuel ratio.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980s and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000s. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort

  18. P2X3 receptors mediate visceral hypersensitivity during acute chemically-induced colitis and in the post-inflammatory phase via different mechanisms of sensitization.

    Directory of Open Access Journals (Sweden)

    Annemie Deiteren

    Full Text Available Experiments using P2X3 knock-out mice or more general P2X receptor antagonists suggest that P2X3 receptors contribute to visceral hypersensitivity. We aimed to investigate the effect of the selective P2X3 antagonist A-317491 on visceral sensitivity under physiological conditions, during acute colitis and in the post-inflammatory phase of colitis.Trinitrobenzene sulphonic-acid colitis was monitored by colonoscopy: on day 3 to confirm the presence of colitis and then every 4 days, starting from day 10, to monitor convalescence and determine the exact timepoint of endoscopic healing in each rat. Visceral sensitivity was assessed by quantifying visceromotor responses to colorectal distension in controls, rats with acute colitis and post-colitis rats. A-317491 was administered 30 min prior to visceral sensitivity testing. Expression of P2X3 receptors (RT-PCR and immunohistochemistry and the intracellular signalling molecules cdk5, csk and CASK (RT-PCR were quantified in colonic tissue and dorsal root ganglia. ATP release in response to colorectal distension was measured by luminiscence.Rats with acute TNBS-colitis displayed significant visceral hypersensitivity that was dose-dependently, but not fully, reversed by A-317491. Hypersenstivity was accompanied by an increased colonic release of ATP. Post-colitis rats also displayed visceral hypersensitivity that was dose-dependently reduced and fully normalized by A-317491 without increased release of ATP. A-317491 did not modify visceral sensitivity in controls. P2X3 mRNA and protein expression in the colon and dorsal root ganglia were similar in control, acute colitis and post-colitis groups, while colonic mRNA expression of cdk5, csk and CASK was increased in the post-colitis group only.These findings indicate that P2X3 receptors are not involved in sensory signaling under physiological conditions whereas they modulate visceral hypersensitivity during acute TNBS-colitis and even more so in the post

  19. A methodology for the evaluation of fuel rod failures under transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.R.; Machiels, A.J. [ANATECH, San Diego, CA (United States)]|[EPRI, Palo Alto (United States)

    2004-07-01

    , with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite element code. The characterization of hydrides-dependent properties of high-burnup fuel cladding is the main fea-ture of this constitutive model. The third element in the overall process is to utilize this material model and its host finite element code in the structural analysis of a transportation cask subjected to bounding accident loading to cal-culate fuel rod failures and failure mode configurations. This requires detailed modeling of the transport cask and its internal structure, which include canister, basket, fuel assembly grids and fuel rods. The overall methodology is described in the paper.

  20. Does the radiation from the interim storage in Gorleben affect the sex ratio of newborn children?; Beeinflusst die Strahlung aus dem Zwischenlager in Gorleben das Geschlechterverhaeltnis von Neugeborenen?

    Energy Technology Data Exchange (ETDEWEB)

    Engelmann, H.W.; Schulze, H.; Wede, S. [GNS Gesellschaft fuer Nuklear-Service mbH, Gorleben (Germany); Mueller, S. [Studsvik GmbH, Pforzheim (Germany)

    2015-07-01

    In the professional world but especially in public, the question is discussed whether ionizing radiation from nuclear facilities has a significant impact on the secondary sex ratio of newborn children in the vicinity of the plants. This issue is of exceptional importance in the region around Gorleben, where the opposition to nuclear facilities and activities for decades is particularly strong. At the site borders of the interim storage facility (TBL-G) of GNS the effective individual dose is about 0.2 mSv per year, mainly caused by neutron irradiation from 108 casks with high-level radioactive waste from reprocessing. In the surrounding villages there is no radiation measurable. Statistical studies allegedly have shown evidence that in some villages in the area and during certain periods, proportionately fewer girls were born in comparison to the average for the Federal Republic of Germany. Based on these purely statistical results henceforward was also alleged that neutron-induced secondary effects such as activation or secondary gamma radiation would be responsible for it. Monte Carlo calculations and special measurements yielded values of the dose at the plant border for activation products less than E-04 mSv/a and for secondary gamma radiation of about E-03 mSv/a. These results indicate that the ionizing radiation from the Gorleben interim storage facility cannot be held accountable for shifts of the secondary sex ratio.

  1. Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

    Energy Technology Data Exchange (ETDEWEB)

    Howden, G.F.

    1994-10-24

    A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions.

  2. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  3. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  4. Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

    Energy Technology Data Exchange (ETDEWEB)

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2015-10-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

  5. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  6. Mitigation of cesium and cobalt contamination on the surfaces of RAM packages

    Energy Technology Data Exchange (ETDEWEB)

    Krumhansl, J.; Bonhomme, F.; McConnell, P. [Sandia National Labs., Albuquerque (United States); Kapoor, A. [U.S. Dept. of Energy, Albuquerque (United States)

    2004-07-01

    Techniques for mitigating the adsorption of {sup 137}Cs and {sup 60}Co on metal surfaces (e.g., RAM packages) exposed to contaminated water (e.g., spent-fuel pools) has been developed and experimentally verified. The techniques are also effective in removing some of the {sup 60}Co and {sup 137}Cs that may have been adsorbed on the surfaces after removal from the contaminated water. The principle for the {sup 137}Cs mitigation technique is based upon ion-exchange processes. In contrast, {sup 60}Co contamination primarily resides in minute particles of CRUD that become lodged on cask surfaces. CRUD is an insoluble Fe-Ni-Cr oxide that forms colloidal-sized particles as stainless steels corrode. Because of the similarity between Ni{sup +2} and Co{sup +2}, CRUD is able to scavenge and retain traces of cobalt as it forms. A number of organic compounds have a great specificity for combining with nickel and cobalt. Ongoing research is investigating the effectiveness of chemical complexing agent, EDTA, with regard to its ability to dissolve the host phase (CRUD) thereby liberating the entrained 60Co into a solution where it can be rinsed away.

  7. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

  8. Review of NDE Methods for Detection and Monitoring of Atmospheric SCC in Welded Canisters for the Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sorenson, Ken B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-01-14

    Dry cask storage systems (DCSSs) for used nuclear fuel (UNF) were originally envisioned for storage periods of short duration (~ a few decades). However, uncertainty challenges the opening of a permanent repository for UNF implying that UNF will need to remain in dry storage for much longer durations than originally envisioned (possibly for centuries). Thus, aging degradation of DCSSs becomes an issue that may not have been sufficiently considered in the design phase and that can challenge the efficacy of very long-term storage of UNF. A particular aging degradation concern is atmospheric stress corrosion cracking (SCC) of DCSSs located in marine environments. In this report, several nondestructive (NDE) methods are evaluated with respect to their potential for effective monitoring of atmospheric SCC in welded canisters of DCSSs. Several of the methods are selected for evaluation based on their usage for in-service inspection applications in the nuclear power industry. The technologies considered include bulk ultrasonic techniques, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves.

  9. Technical changes that would contribute to success in the civilian radioactive waste management program

    Energy Technology Data Exchange (ETDEWEB)

    Ramspott, L.D.

    1993-10-01

    Many changes have taken place since the SCP safety strategy was formulated; it needs to be revised or replaced. Four concepts would aid in the shift from a rigid, ecelctic, schedule-driven, all-or-nothing program to an incremental, evolving, and experimental but integrated program. These are a simple safety case, reversability, demonstrability, and decoupling operations of a repository from operation of reactors. A simple safety case based on containment can be made for a repository at Yucca Mountain. This containment strategy is based on the dryness of openings at Yucca Mountain, Extended Dry heat management, and long-lived containers. Reversibility is technically believable at Yucca Mountain because of extended retrievability and drift emplacement, if an MRS were co-located with the repository. Because the rock is unsaturated, extended retrievability is technically feasible at Yucca Mountain. Demonstrability could be improved at Yucca Mountain by planning for incremental progression toward operation and closure of a repository, possibly including a shift to underground retrievable storage. Demonstrability can also be improved by using natural analogs. Repository operation can be decoupled from reactor operation by use of an unconstrained MRS facility or at-reactor dry storage and multipurpose storage canister/casks.

  10. Proceedings of the Chornobyl phytoremediation and biomass energy conversion workshop

    Energy Technology Data Exchange (ETDEWEB)

    Hartley, J. [Pacific Northwest National Lab., Richland, WA (United States); Tokarevsky, V. [State Co. for Treatment and Disposal of Mixed Hazardous Waste (Ukraine)

    1998-06-01

    Many concepts, systems, technical approaches, technologies, ideas, agreements, and disagreements were vigorously discussed during the course of the 2-day workshop. The workshop was successful in generating intensive discussions on the merits of the proposed concept that includes removal of radionuclides by plants and trees (phytoremediation) to clean up soil in the Chornobyl Exclusion Zone (CEZ), use of the resultant biomass (plants and trees) to generate electrical power, and incorporation of ash in concrete casks to be used as storage containers in a licensed repository for low-level waste. Twelve years after the Chornobyl Nuclear Power Plant (ChNPP) Unit 4 accident, which occurred on April 26, 1986, the primary 4radioactive contamination of concern is from radioactive cesium ({sup 137}Cs) and strontium ({sup 90}Sr). The {sup 137}Cs and {sup 90}Sr were widely distributed throughout the CEZ. The attendees from Ukraine, Russia, Belarus, Denmark and the US provided information, discussed and debated the following issues considerably: distribution and characteristics of radionuclides in CEZ; efficacy of using trees and plants to extract radioactive cesium (Cs) and strontium (Sr) from contaminated soil; selection of energy conversion systems and technologies; necessary infrastructure for biomass harvesting, handling, transportation, and energy conversion; radioactive ash and emission management; occupational health and safety concerns for the personnel involved in this work; and economics. The attendees concluded that the overall concept has technical and possibly economic merits. However, many issues (technical, economic, risk) remain to be resolved before a viable commercial-scale implementation could take place.

  11. Analysis of the risk of transporting spent nuclear fuel by train

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  12. 30 years of experience in safe transportation of nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, K. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan)

    2004-07-01

    In April 2003, Nuclear Fuel Transport Co., Ltd. (NFT) marked the 30{sup th} anniversary of its founding. NFT was established in 1973 and in 1978, commenced SF transport to the reprocessing plant in Tokai-mura. And then, after making preparations to transport nuclear materials to the various facilities at the Nuclear Fuel Cycle Center in Rokkasho-mura, NFT successfully started transportation of LLW (low level waste) to Rokksho-mura's LLW disposal center in 1992, domestic land transportation of HLW returned from overseas to the HLW storage center in 1995, domestic land transportation of natural hexafluoride delivered from overseas to the uranium enrichment plant in 1996, and transportation of SF to the reprocessing plant in 2000. NFT has realized an annual SF transportation capacity of 300 MTU and is currently making great company wide efforts to meet the Rokkasho Reprocessing Plant's future SF annual reprocessing capacity of 800MTU. At the end of FY2003, NFT had successfully transported 560 casks (about 1,730 MTU) of SF in more than 200 voyages in total, about 160,000 drums of LLW in around 100 voyages in total. This paper introduces the record of safe transport and its experience over the past 30 years and prospect for future transport business.

  13. Study on containerisation of irradiated fuel at JRC ISPRA for medium/long-term storage

    Energy Technology Data Exchange (ETDEWEB)

    Bertelli, S.; Bielli, G.; Covini, R.

    2001-07-01

    During the last 40 years big amounts of wastes arising from past experiments have been generated at JRC Ispra. These wastes are now stored on site in unconditioned form and must be characterised and re-conditioned to ensure their acceptance by future repositories. Among the several types of wastes produced, spent fuel has a great importance in JRC waste management. In fact, there are more than the tons of irradiated material, varying widely from commercial to experimental fuel elements or pins, in form of oxides and metal fuel, with very different geometry, dry and wet stored. The biggest part of it can be considered as unirradiated, according to the IAEA regulations (ST-1, 1996), while a relatively small amount (nearly 720 kg U total) are to be considered as irradiated and treated accordingly. Studies on the most suitable solutions for medium/long term storage for such irradiated experimental fuel are being performed, taking into account two main options, containerisation of the nuclear materials, as it is, in suitable casks or reprocessing. The technical aspects of the project for containerisation are here discussed. (Author)

  14. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  15. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  16. Efficient Reconstruction of Heterogeneous Networks from Time Series via Compressed Sensing.

    Directory of Open Access Journals (Sweden)

    Long Ma

    Full Text Available Recent years have witnessed a rapid development of network reconstruction approaches, especially for a series of methods based on compressed sensing. Although compressed-sensing based methods require much less data than conventional approaches, the compressed sensing for reconstructing heterogeneous networks has not been fully exploited because of hubs. Hub neighbors require much more data to be inferred than small-degree nodes, inducing a cask effect for the reconstruction of heterogeneous networks. Here, a conflict-based method is proposed to overcome the cast effect to considerably reduce data amounts for achieving accurate reconstruction. Moreover, an element elimination method is presented to use the partially available structural information to reduce data requirements. The integration of both methods can further improve the reconstruction performance than separately using each technique. These methods are validated by exploring two evolutionary games taking place in scale-free networks, where individual information is accessible and an attempt to decode the network structure from measurable data is made. The results demonstrate that for all of the cases, much data are saved compared to that in the absence of these two methods. Due to the prevalence of heterogeneous networks in nature and society and the high cost of data acquisition in large-scale networks, these approaches have wide applications in many fields and are valuable for understanding and controlling the collective dynamics of a variety of heterogeneous networked systems.

  17. Efficient Reconstruction of Heterogeneous Networks from Time Series via Compressed Sensing.

    Science.gov (United States)

    Ma, Long; Han, Xiao; Shen, Zhesi; Wang, Wen-Xu; Di, Zengru

    2015-01-01

    Recent years have witnessed a rapid development of network reconstruction approaches, especially for a series of methods based on compressed sensing. Although compressed-sensing based methods require much less data than conventional approaches, the compressed sensing for reconstructing heterogeneous networks has not been fully exploited because of hubs. Hub neighbors require much more data to be inferred than small-degree nodes, inducing a cask effect for the reconstruction of heterogeneous networks. Here, a conflict-based method is proposed to overcome the cast effect to considerably reduce data amounts for achieving accurate reconstruction. Moreover, an element elimination method is presented to use the partially available structural information to reduce data requirements. The integration of both methods can further improve the reconstruction performance than separately using each technique. These methods are validated by exploring two evolutionary games taking place in scale-free networks, where individual information is accessible and an attempt to decode the network structure from measurable data is made. The results demonstrate that for all of the cases, much data are saved compared to that in the absence of these two methods. Due to the prevalence of heterogeneous networks in nature and society and the high cost of data acquisition in large-scale networks, these approaches have wide applications in many fields and are valuable for understanding and controlling the collective dynamics of a variety of heterogeneous networked systems.

  18. Design Report of Volume Reduction and Handling Equipment of ACPF Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, K. C.; Ku, J. H.; Lee, W. K.; Lee, E. P.; Choung, W. M.; Cho, I. J.; Kook, D. H.; You, G. S.; Youn, J. S

    2007-01-15

    For the efficient disposal of wastes from the ACPF (Advanced spent fuel Conditioning Process Facility), hot cell filter press equipment, vacuum press equipment of compatible wastes and hot cell auxiliary crane were developed. The filter press equipment installed in hot cell is pneumatically driven and remotely operated. In order to avoid the contamination of hot cell by dust particles from filters being compressed, it is designed to compress filters with them put into the basket of the solid wastes cask. The performance test showed that the height of a filter was reduced by more than 75% and a basket could contain up to four compressed filters. The vacuum press equipment reduces the volume of the plastic bag containing compatible wastes such as cloth, plastic film and paper by vacuumizing the bag. It consists of the vacuum pump, the suction and HEPA filters and the manual pneumatic valve, and is designed to be easily handled and managed. Through the performance test, it is observed that the volume of wastes was reduced to a third. Also an additional crane of 300 kg capacity was manufactured and installed in the ACP hot cell to handle the salt ingot wastes container located where the overhead crane is not accessible. This report presenting the details of the design, manufacturing and performance of these equipment will be used as technical materials for the operation of the facility and the development of equipment.

  19. Cobalt-60 heat source demonstration program. Phase III. Fabrication. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1973-06-01

    Significant accomplishments completed during Phase III of the /sup 60/Co Heat Source Demonstration program include the following: encapsulation of 2 MCi of /sup 60/Co; fabrication of the heat source, including the ASME coded pressure vessel/core assembly, and biological shielding; endurance testing of a prototype heat pipe for a period of 28 months; fabrication and qualification of the heat pipe emergency cooling subsystem; issue of the safety evaluation report, reference 3, and the operations manual, reference 4; and heat source assembly. The planned demonstration test program was modified to include testing of a total power system. Based on an evaluation of available power conversion systems, which included the closed-cycle Brayton and organic Rankine systems, the closed-cycle Brayton system was selected for use. Selection was based on advantages offered by the direct coupling of this conversion system with the gas-cooled heat source. In implementing the test program, the AiResearch BCD power conversion system was to be coupled to the heat source following initial heat source performance testing and part way through the endurance test. In accordance with the program redirection the following Phase IV checkout operations were completed to evaluate procedural and hardware acceptability: heat source dummy fueling; fueling cask sielding survey; and heat source shielding survey (single pin). Completion of these latter activities verified the acceptability of critical characteristics of the heat source and its supporting equipment.

  20. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  1. Underground Corrosion of Activated Metals in an Arid Vadose Zone Environment

    Energy Technology Data Exchange (ETDEWEB)

    Adler Flitton, Mariana Kay; Mizia, Ronald Eugene; Bishop, Carolyn Wagoner

    2002-04-01

    The subsurface radioactive disposal site located at the Idaho National Engineering and Environmental Laboratory contains neutron-activated metals from nonfuel nuclear-reactor- core components. A long-term corrosion test is being conducted to obtain site-specific corrosion rates to support efforts to more accurately estimate the transfer of activated elements in an arid vadose zone environment. The tests use nonradioactive metal coupons representing the prominent neutron-activated material buried at the disposal location, namely, Type 304L stainless steel, Type 315L stainless steel, nickel-chromium alloy (UNS NO7718), beryllium, aluminum 6061-T6, and a zirconium alloy, (UNS R60804). In addition, carbon steel (the material presently used in the cask disposal liners and other disposal containers) and a duplex stainless steel (UNS S32550) (the proposed material for the high- integrity disposal containers) are also included in the test program. This paper briefly describes the test program and presents the early corrosion rate results after 1 year and 3 years of underground exposure.

  2. Underground Corrosion of Activated Metals, 6-Year Exposure Analysis

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Adler Flitton; T. S. Yoder

    2006-03-01

    The subsurface radioactive disposal site located at the Idaho National Laboratory contains neutronactivated metals from non-fuel nuclear-reactor-core components. A long-term underground corrosion test is being conducted to obtain site-specific corrosion rates to support efforts to more accurately estimate the transfer of activated elements in the surrounding arid vadose zone environment. The test uses nonradioactive metal coupons representing the prominent neutron-activated materials buried at the disposal location, namely, Type 304L stainless steel (UNS S30403), Type 316L stainless steel (S31603), nickel-chromium alloy (UNS NO7718), beryllium, aluminum 6061-T6 (A96061), and a zirconium alloy (UNS R60804). In addition, carbon steel (the material presently used in the cask disposal liners and other disposal containers) and a duplex stainless steel (UNS S32550) are also included in the test. This paper briefly describes the ongoing test and presents the results of corrosion analysis from coupons exposed underground for 1, 3, and 6 years.

  3. Underground Corrosion of Activated Metals in an Arid Vadose Zone Environment

    Energy Technology Data Exchange (ETDEWEB)

    Adler Flitton, M.K; Mizia, R.E.; Bishop, C.W.

    2001-10-24

    The subsurface radioactive disposal site located at the Idaho National Engineering and Environmental Laboratory contains neutron-activated metals from nonfuel nuclear-reactor- core components. A long-term corrosion test is being conducted to obtain site-specific corrosion rates to support efforts to more accurately estimate the transfer of activated elements in an arid vadose zone environment. The tests use nonradioactive metal coupons representing the prominent neutron-activated material buried at the disposal location, namely, Type 304L stainless steel, Type 315L stainless steel, nickel-chromium alloy (UNS NO7718), beryllium, aluminum 6061-T6, and a zirconium alloy, (UNS R60804). In addition, carbon steel (the material presently used in the cask disposal liners and other disposal containers) and a duplex stainless steel (UNS S32550) (the proposed material for the high- integrity disposal containers) are also included in the test program. This paper briefly describes the test program and presents the early corrosion rate results after 1 year and 3 years of underground exposure.

  4. Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

    Energy Technology Data Exchange (ETDEWEB)

    Triplett, C.E.

    1996-12-01

    This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.

  5. HFIR spent fuel management alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-10-15

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

  6. Evaluation of Storage for Transportation Equipment, Unfueled Convertors, and Fueled Convertors at the INL for the Radioisotope Power Systems Program

    Energy Technology Data Exchange (ETDEWEB)

    S. G. Johnson; K. L. Lively

    2010-05-01

    This report contains an evaluation of the storage conditions required for several key components and/or systems of the Radioisotope Power Systems (RPS) Program at the Idaho National Laboratory (INL). These components/systems (transportation equipment, i.e., type ‘B’ shipping casks and the radioisotope thermo-electric generator transportation systems (RTGTS), the unfueled convertors, i.e., multi-hundred watt (MHW) and general purpose heat source (GPHS) RTGs, and fueled convertors of several types) are currently stored in several facilities at the Materials and Fuels Complex (MFC) site. For various reasons related to competing missions, inherent growth of the RPS mission at the INL and enhanced efficiency, it is necessary to evaluate their current storage situation and recommend the approach that should be pursued going forward for storage of these vital RPS components and systems. The reasons that drive this evaluation include, but are not limited to the following: 1) conflict with other missions at the INL of higher priority, 2) increasing demands from the INL RPS Program that exceed the physical capacity of the current storage areas and 3) the ability to enhance our current capability to care for our equipment, decrease maintenance costs and increase the readiness posture of the systems.

  7. Synchronous fluorescence spectroscopy for analysis of wine and wine distillates

    Science.gov (United States)

    Andreeva, Ya.; Borisova, E.; Genova, Ts.; Zhelyazkova, Al.; Avramov, L.

    2015-01-01

    Wine and brandies are multicomponent systems and conventional fluorescence techniques, relying on recording of single emission or excitation spectra, are often insufficient. In such cases synchronous fluorescence spectra can be used for revealing the potential of the fluorescence techniques. The technique is based on simultaneously scanning of the excitation and emission wavelength with constant difference (Δλ) maintained between them. In this study the measurements were made using FluoroLog3 spectrofluorimeter (HORIBA Jobin Yvon, France) and collected for excitation and emission in the wavelength region 220 - 700 nm using wavelength interval Δλ from 10 to 100 nm in 10 nm steps. This research includes the results obtained for brandy and red wine samples. Fluorescence analysis takes advantage in the presence of natural fluorophores in wines and brandies, such as gallic, vanillic, p-coumaric, syringic, ferulic acid, umbelliferone, scopoletin and etc. Applying of synchronous fluorescence spectroscopy for analysis of these types of alcohols allows us to estimate the quality of wines and also to detect adulteration of brandies like adding of a caramel to wine distillates for imitating the quality of the original product aged in oak casks.

  8. Design of spent-fuel concrete pit dry storage and handling system

    Energy Technology Data Exchange (ETDEWEB)

    Tamaki, H.; Natsume, T.; Maruoka, K.; Yokoyama, T. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan)

    1998-07-01

    An advanced dry storage system design with highly improved storage efficiency of spent nuclear fuel has been developed. The new concept 'Concrete Pit Dry Storage System' realizes a safe and economical solution to an increasing demand of storing spent fuel assemblies (SFAs) generated from commercial nuclear power reactors. The system is basically composed of a large mass concrete module which has densely arranged pit boreholes, sealed canisters containing spent fuel assemblies and a canister handling system. The system is characterized by the following advantages compared with the existing concrete module type storage systems: higher storage efficiency can be achieved by the storage module filled with concrete which also gives a high shielding performance; simple handling technology is used for transfer and installation of the canisters at the storage facility as well as the transport cask of the canisters, surface contamination of the canister is prevented; lower radiation around the storage area is provided to reduce radiation exposure during handling and storage; high structural integrity of the facility is maintained by the concrete module with a simple construction ; the ventilation gallery introducing cooling air air to the bit borehole has an enough draft height to improve cooling performance of the system; a result of the design concept, the storage system can store higher burn-up SFAs with a short cooling period. (authors)

  9. FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: ``Super'' alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

    Science.gov (United States)

    O'Grady, Joseph; Bument, Arlden; Siegel, Edward

    2011-03-01

    Carbides solid-state chemistry domination of old/new nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines is austenitic/FCC Ni/Fe-based (so miscalled)"super"alloys(182/82;Hastelloy-X,600,304/304L-SSs,...690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-disease(WD) [J.Appl.Phys.17,857 (46)]/Ostwald-ripening/spinodal-decomposition/overageing-embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google: fLeaksCouldKill > ; - Siegel [ J . Mag . Mag . Mtls . 7 , 312 (78) = atflickr . comsearchonGiant - Magnotoresistance [Fert" [PRL(1988)]-"Gruenberg"[PRL(1989)] 2007-Nobel]necessitating NRC inspections on 40+25=65 Westin"KL"ouse PWRs(12/2006)]-Lai [Met.Trans.AIME, 9A,827(78)]-Sabol-Stickler[Phys.Stat.Sol.(70)]-Ashpahani[ Intl.Conf. Hydrogen in Metals, Paris(1977]-Russell [Prog.Mtls.Sci.(1983)]-Pollard [last UCS rept.(9/1995)]-Lofaro [BNL/DOE/NRC Repts.]-Pringle [ Nuclear-Power:From Physics to Politics(1979)]-Hoffman [animatedsoftware.com], what DOE/NRC MISlabels as "butt-welds" "stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrittlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n"u"tional-la"v"atories sabotage!!!

  10. FRAUD/SABOTAGE Killing Nuclear-Reactors!!! ``Super"alloys GENERIC ENDEMIC Wigner's-Disease IN-stability!!!

    Science.gov (United States)

    Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

    2010-03-01

    Siegel [[J.Mag.Mag.Mtls.7,312(78); PSS(a)11,45(72); Semis.& Insuls.5(79)] (at: ORNL, ANS, Westin``KL"ouse, PSEG, IAEA, ABB) warning of old/new nuclear-reactors/spent-fuel-casks/refineries/ jet/missile/rocket-engines austenitic/FCC Ni/Fe-based (so MIS- called)``super"alloys(182/82;Hastelloy-X; 600;304/304L-SSs; 690 !!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's- diseas(WD)[J.Appl.Phys.17,857(46)]; Ostwald-ripening; spinodal- decomposition; overageing-embrittlement; thermomechanical- INstability: Mayo[Google: ``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: [Siegel<<<``Fert"(88) 2007-Nobel/Wolf/Japan-prizes]necessitating NRC inspections on 40+25=65 Westin``KL"ouse PWRs(12/06)]; Lai[Met.Trans.AIME,9A,827 (78)]-Sabol-Stickler[PSS(70)]; Ashpahani[Intl.Conf. H in Metals (77)]; Russell[Prog. Mtls.Sci.(83)]; Pollard[last UCS rept. (9/95)]; Lofaro[BNL/DOE/NRC Repts.]; Pringle[Nuclear-Power:From Physics to Politics(79)]; Hoffman[animatedsoftware.com],...what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrit- tlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n``u''tional-la``v''atories sabotage!!!

  11. FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: "Super"alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

    Science.gov (United States)

    Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

    2010-03-01

    Carbides solid-state chemistry domination of old/new nuclear- reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe-based(so miscalled)``super"alloys(182/82; Hastelloy-X,600,304/304L-SSs,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-diseas(WD)[J.Appl.Phys.17,857 (1946)]/Ostwald-ripening/spinodal-decomposition/overageing- embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google:``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: Siegel[J.Mag.Mag.Mtls.7,312 (1978)]<<<``Fert"-"Gruenberg"(1988/89)2007-physics Nobel/Wolf/ Japan-prizes]necessitating NRC-inspections of 40+25 = 65 Westin- ``KLouse PWRs(12/2006)]-Lai[Met.Trans.AIME,9A,827(1978)]-Sabol- Stickler[Phys.Stat.Sol.(1970)]-Ashpahani[Intl.Conf. H in Metals, Paris(1977]-Russell[Prog.Mtls.Sci.(1983)]-Pollard[last UCS rept. (9/1995)]-Lofaro[BNL/DOE/NRC Repts.]-Pringle[Nuclear-Power:From Physics to Politics(1979)]-Hoffman[animatedsoftware.com], what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embritt- lement caused brittle-fracture cracking from early/ongoing AEC/ DOE-n"u"tional-la"v"atories sabotage!!!

  12. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  13. Neuronal cell adhesion genes: Key players in risk for schizophrenia, bipolar disorder and other neurodevelopmental brain disorders?

    Science.gov (United States)

    Corvin, Aiden P

    2010-01-01

    The major mental disorders, schizophrenia and bipolar disorder are substantially heritable. Recent genomic studies have identified a small number of common and rare risk genes contributing to both disorders and support epidemiological evidence that genetic susceptibility overlaps between them. Prompted by the question of whether risk genes cluster in specific molecular pathways or implicate discrete mechanisms we and others have developed hypothesis-free methods of investigating genome-wide association datasets at a pathway-level. The application of our method to the 212 experimentally-derived pathways in the Kyoto Encycolpaedia of Genes and Genomes (KEGG) database identified significant association between the cell adhesion molecule (CAM) pathway and both schizophrenia and bipolar disorder susceptibility across three GWAS datasets. Interestingly, a similar approach applied to an autistic spectrum disorders (ASDs) sample identified a similar pathway and involved many of the same genes. Disruption of a number of these genes (including NRXN1, CNTNAP2 and CASK) are known to cause diverse neurodevelopmental brain disorder phenotypes including schizophenia, autism, learning disability and specific language disorder. Taken together these studies bring the CAM pathway sharply into focus for more comprehensive DNA sequencing to identify the critical genes, and investigate their relationships and interaction with environmental risk factors in the expression of many seemingly different neurodevelopmental disorders.

  14. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    NA

    2005-10-25

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

  15. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  16. Analytical Chemistry Laboratory (ACL) procedure compendium. Volume 2, Sample preparation methods

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    This volume contains the interim change notice for sample preparation methods. Covered are: acid digestion for metals analysis, fusion of Hanford tank waste solids, water leach of sludges/soils/other solids, extraction procedure toxicity (simulate leach in landfill), sample preparation for gamma spectroscopy, acid digestion for radiochemical analysis, leach preparation of solids for free cyanide analysis, aqueous leach of solids for anion analysis, microwave digestion of glasses and slurries for ICP/MS, toxicity characteristic leaching extraction for inorganics, leach/dissolution of activated metal for radiochemical analysis, extraction of single-shell tank (SST) samples for semi-VOC analysis, preparation and cleanup of hydrocarbon- containing samples for VOC and semi-VOC analysis, receiving of waste tank samples in onsite transfer cask, receipt and inspection of SST samples, receipt and extrusion of core samples at 325A shielded facility, cleaning and shipping of waste tank samplers, homogenization of solutions/slurries/sludges, and test sample preparation for bioassay quality control program.

  17. The microbial diversity of an industrially produced lambic beer shares members of a traditionally produced one and reveals a core microbiota for lambic beer fermentation.

    Science.gov (United States)

    Spitaels, Freek; Wieme, Anneleen D; Janssens, Maarten; Aerts, Maarten; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

    2015-08-01

    The microbiota involved in lambic beer fermentations in an industrial brewery in West-Flanders, Belgium, was determined through culture-dependent and culture-independent techniques. More than 1300 bacterial and yeast isolates from 13 samples collected during a one-year fermentation process were identified using matrix-assisted laser desorption/ionization time-of-flight mass spectrometry followed by sequence analysis of rRNA and various protein-encoding genes. The bacterial and yeast communities of the same samples were further analyzed using denaturing gradient gel electrophoresis of PCR-amplified V3 regions of the 16S rRNA genes and D1/D2 regions of the 26S rRNA genes, respectively. In contrast to traditional lambic beer fermentations, there was no Enterobacteriaceae phase and a larger variety of acetic acid bacteria were found in industrial lambic beer fermentations. Like in traditional lambic beer fermentations, Saccharomyces cerevisiae, Saccharomyces pastorianus, Dekkera bruxellensis and Pediococcus damnosus were the microorganisms responsible for the main fermentation and maturation phases. These microorganisms originated most probably from the wood of the casks and were considered as the core microbiota of lambic beer fermentations.

  18. Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Mo [Univ. of California, Irvine, CA (United States); Nakshatrala, Kalyana [Univ. of Houston, TX (United States); William, Kasper [Univ. of Houston, TX (United States); Xi, Yungping [Univ. of Colorado, Boulder, CO (United States)

    2017-02-08

    The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

  19. A Review of Removable Surface Contamination on Radioactive Materials Transportation Containers

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Jr, W. E.; Watson, E. C.; Murphy, D. W.; Harrer, B. J.; Harty, R.; Aldrich, J. M.

    1981-05-01

    This report contains the results of a study sponsored by the U.S. Nuclear Regulatory Commission (NRC) of removable surface contamination on radioactive materials transportation containers. The purpose of the study is to provide information to the NRC during their review of existing regulations. Data was obtained from both industry and literature on three major topics: 1) radiation doses, 2) economic costs, and 3) contamination frequencies. Containers for four categories of radioactive materials are considered including radiopharmaceuticals, industrial sources, nuclear fuel cycle materials, and low-level radioactive waste. Assumptions made in this study use current information to obtain realistic yet conservative estimates of radiation dose and economic costs. Collective and individual radiation doses are presented for each container category on a per container basis. Total doses, to workers and the public, are also presented for spent fuel cask and low-level waste drum decontamination. Estimates of the additional economic costs incurred by lowering current limits by factors of 10 and 100 are presented. Current contamination levels for each category of container are estimated from the data collected. The information contained in this report is designed to be useful to the NRC in preparing their recommendations for new regulations.

  20. Greater-Than-Class C Low-Level Radioactive Waste Transportation Strategy report and institutional plan

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, R.C.; Tyacke, M.J.

    1995-01-01

    This document contains two parts. Part I, Greater-Than-Class-C Low-Level Radioactive Waste Transportation Strategy, addresses the requirements, responsibilities, and strategy to transport and receive these wastes. The strategy covers (a) transportation packaging, which includes shipping casks and waste containers; (b) transportation operations relating to the five facilities involved in transportation, i.e., waste originator, interim storage, dedicated storage, treatment, and disposal; (c) system safety and risk analysis; (d) routes; (e) emergency preparedness and response; and (o safeguards and security. A summary of strategic actions is provided at the conclusion of Part 1. Part II, Institutional Plan for Greater-Than-Class C Low-Level Radioactive Waste Packaging and Transportation, addresses the assumptions, requirements, and institutional plan elements and actions. As documented in the Strategy and Institutional Plan, the most challenging issues facing the GTCC LLW Program shipping campaign are institutional issues closely related to the strategy. How the Program addresses those issues and demonstrates to the states, local governments, and private citizens that the shipments can and will be made safely will strongly affect the success or failure of the campaign.