WorldWideScience

Sample records for dry-storage dual-purpose casks

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. Activation analysis of dual-purpose metal cask after the end of design lifetime for decommission

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Man; Ku, Ji Young; Dho Ho Seog; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Ko, Jae Hun [Korea Nuclear Engineering and Service Co., Daejeon (Korea, Republic of)

    2016-12-15

    The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5·ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, {sup 60}Co had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as {sup 28}Al and {sup 24}Na had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that

  7. 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.

  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. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  10. 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.

  11. 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.

  12. 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

  13. 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

  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. 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

  16. 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...

  17. 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.

  18. 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.

  19. 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

  20. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

    Science.gov (United States)

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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)

  9. 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...

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. A Well Established System For The Dry Storage Of Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Skrzyppek, Juergen; Kim, Josef Du-III [SMART Power Company, Essen (Germany)

    2015-05-15

    The German company GNS Gesellschaft fur Nuklear-Service mbH today looks back on more than 30 years of operational experience with dual-purpose casks. Following customer demands, GNS developed two different cask types for SNF the CASTOR and the CONSTOR cask type. While the CASTOR type is optimized for high thermal loads which allows loading after extremely short cooling times and/or high burn-up of the SNF the CONSTOR type is cost optimized for the cost-efficient storage of large quantities of cooler SNF. By now almost 1,300 GNS-casks are in operation worldwide. In Germany alone, more than 1000 CASTOR casks are stored with individual storage periods of up to 30 years. Taking into account the additional casks that have to be manufactured, loaded and stored during the final years of the German Nuclear Phase-Out, there will be 2000 casks by GNS in operation worldwide. The presentation will give an overview over several national and international projects and show the bandwidth of customized solutions by GNS.

  16. Probabilistic risk assessment of aircraft impact on a spent nuclear fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Almomani, Belal, E-mail: balmomani@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Sanghoon, E-mail: shlee1222@kmu.ac.kr [Department of Mechanical and Automotive Engineering, Keimyung University, Dalgubeol-daero 1095, Dalseo-gu, Daegu (Korea, Republic of); Jang, Dongchan, E-mail: dongchan.jang@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Kang, Hyun Gook, E-mail: kangh6@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY 12180 (United States)

    2017-01-15

    Highlights: • A new risk assessment frame is proposed for aircraft impact into an interim dry storage. • It uses event tree analysis, response-structural analysis, consequence analysis, and Monte Carlo simulation. • A case study of the proposed procedure is presented to illustrate the methodology’s application. - Abstract: This paper proposes a systematic risk evaluation framework for one of the most significant impact events on an interim dry storage facility, an aircraft crash, by using a probabilistic approach. A realistic case study that includes a specific cask model and selected impact conditions is performed to demonstrate the practical applicability of the proposed framework. An event tree analysis of an occurred aircraft crash that defines a set of impact conditions and storage cask response is constructed. The Monte-Carlo simulation is employed for the probabilistic approach in consideration of sources of uncertainty associated with the impact loads onto the internal storage casks. The parameters for representing uncertainties that are managed probabilistically include the aircraft impact velocity, the compressive strength of the reinforced concrete wall, the missile shape factor, and the facility wall thickness. Failure probabilities of the impacted wall and a single storage cask under direct mechanical impact load caused by the aircraft crash are estimated. A finite element analysis is applied to simulate the postulated direct engine impact load onto the cask body, and a source term analysis for associated releases of radioactive materials as well as an off-site consequence analysis are performed. Finally, conditional risk contribution calculations are represented by an event tree model. Case study results indicate that no severe risk is presented, as the radiological consequences do not exceed regulatory exposure limits to the public. This risk model can be used with any other representative detailed parameters and reference design concepts for

  17. 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''...

  18. 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

  19. 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)

  20. 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.

  1. 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).

  2. CASTOR {sup ®} and CONSTOR {sup ®}. A well established system for the dry storage of spent fuel and high level waste

    Energy Technology Data Exchange (ETDEWEB)

    Wimmer, Hannes; Skrzyppek, Juergen; Koebl, Michael [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2015-06-01

    The German company GNS Gesellschaft fuer Nuklear-Service mbH today looks back on more than 30 years of operational experience with dual-purpose casks for the transport and storage of spent nuclear fuel (SNF) from nuclear power plants and high level waste (HLW) from reprocessing. Following customer demands, GNS developed two different cask types for SNF. By now, almost 1,300 GNS-casks are in operation worldwide. This article gives an overview over several national and international projects and shows the bandwidth of customised solutions by GNS.

  3. 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

  4. Evolution of spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Standring, Paul Nicholas [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Fuel Cycle and Waste Technology; Takats, Ferenc [TS ENERCON KFT, Budapest (Hungary)

    2016-11-15

    Around 10,000 tHM of spent fuel is discharged per year from the nuclear power plants in operation. Whilst the bulk of spent fuel is still held in at reactor pools, 24 countries have developed storage facilities; either on the reactor site or away from the reactor site. Of the 146 operational AFR storage facilities about 80 % employ dry storage; the majority being deployed over the last 20 years. This reflects both the development of dry storage technology as well as changes in politics and trading relationships that have affected spent fuel management policies. The paper describes the various approaches to the back-end of the nuclear fuel cycle for power reactor fuels and provides data on deployed storage technologies.

  5. Summary Report for Capsule Dry Storage Project

    Energy Technology Data Exchange (ETDEWEB)

    JOSEPHSON, W S

    2003-09-04

    There are 1.936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project (CDSP) is conducted under the assumption the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event vitrification of the capsule contents is pursued. A cut away drawing of a typical cesium chloride (CsCI) capsule and the capsule property and geometry information are provided in Figure 1.1. Strontium fluoride (SrF{sub 2}) capsules are similar in design to CsCl capsules. Further details of capsule design, current state, and reference information are given later in this report and its references. Capsule production and life history is covered in WMP-16938, Capsule Characterization Report for Capsule Dry Storage Project, and is briefly summarized in Section 5.2 of this report.

  6. 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 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 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 above-ground configuration.

  7. 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

  8. 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.

  9. 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)

  10. 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.

  11. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  12. Thermal Analysis of a Dry Storage Concept for Capsule Dry Storage Project

    Energy Technology Data Exchange (ETDEWEB)

    JOSEPHSON, W S

    2003-09-04

    There are 1,936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project is conducted under the assumption that the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event that vitrification of the capsule contents is pursued. The Capsule Advisory Panel (CAP) was created by the Project Manager for the Hanford Site Capsule Dry Storage Project (CDSP). The purpose of the CAP is to provide specific technical input to the CDSP; to identify design requirements; to ensure design requirements for the project are conservative and defensible; to identify and resolve emerging, critical technical issues, as requested; and to support technical reviews performed by regulatory organizations, as requested. The CAP will develop supporting and summary documents that can be used as part of the technical and safety bases for the CDSP. The purpose of capsule dry storage thermal analysis is to: (1) Summarize the pertinent thermal design requirements sent to vendors, (2) Summarize and address the assumptions that underlie those design requirements, (3) Demonstrate that an acceptable design exists that satisfies the requirements, (4) Identify key design features and phenomena that promote or impede design success, (5) Support other CAP analyses such as corrosion and integrity evaluations, and (6) Support the assessment of proposed designs. It is not the purpose of this report to optimize or fully analyze variations of postulated acceptable designs. The present evaluation will indicate the impact of various possible design features, but not systematically pursue design improvements obtainable through analysis

  13. Structural analysis of a metal spent-fuel storage cask in an aircraft crash for risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Almomani, Belal, E-mail: balmomani@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Sanghoon, E-mail: shlee1222@kmu.ac.kr [Department of Mechanical and Automotive Engineering, Keimyung University, Dalgubeol-daero 1095, Dalseo-gu, Daegu (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2016-11-15

    Highlights: • Several engine-applied loads with different locations of impact on the storage cask body were implemented. • Cask structural responses due to the influence of engine impact loadings were analyzed. • Leakage path areas from lid closure openings were numerically calculated. • Release fractions that depend on the generated seal opening areas and fuel damage ratios were estimated. - Abstract: Evaluations of the impact resistance of a dry storage cask under mechanical impact loadings resulting from a large commercial aircraft crash have become an important issue for designers and evaluators, in order to promote interim dry storage activities and to evaluate design safety margins. This study presents a method to evaluate the structural integrity of a generic metal cask subjected to various mechanical loading conditions, which represent aircraft engine impacts, on different locations of the cask body. Thirty representative impact conditions are analyzed to provide a comprehensive evaluation of cask damage response. The applied engine impact load–time functions were carefully re-derived by utilizing CRIEPI’s proposed curve through Riera’s approach for six impact velocities, and applied to five locations on a freestanding cask: lateral impacts on the lower half, center of gravity, and upper half of the cask body, corner impact on the lid closure, and vertical impact on the center of the lid closure. A nonlinear dynamic finite element analysis is performed to evaluate the dynamic response of the cask lid closure system and to calculate the lid gaps. The release fractions from the cask to the environment for each impact condition are preliminarily estimated by referring to a proposed methodology from literature. It is believed that this paper presents a systematic process to connect the mechanical analysis of a cask response at the moment of aircraft engine impact with its radiological consequence analysis.

  14. Evaluating the feasibility of new surveillance concept for Dry Storage System through CFD methodology

    Energy Technology Data Exchange (ETDEWEB)

    Tseng, Y.S., E-mail: yungshintseng@gmail.com [Nuclear Science and Technology Development Center, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325, Taiwan, ROC (China); Lin, C.H., E-mail: gp6ej3@gmail.com [Center for Energy and Environmental Research, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325, Taiwan, ROC (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325, Taiwan, ROC (China); Wang, J.R., E-mail: jongrongwang@gmail.com [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Rd., Hsingchu 30013, 325, Taiwan, ROC (China)

    2016-08-01

    Highlights: • Thermal hydraulic behavior of a 3-D dry cask under several off-normal conditions has been numerically investigated by ANSYS/FLUENT. • The simulation methodology was fully validated by comparing the measured results of VSC-17. • The results indicated that many design bases accidents can be early detected by the purposed surveillance method. • A simply determine rule has been developed for future application for dry storage ​monitoring. - Abstract: Since the Dry Storage System (DSS) has passed into a widespread middle-term storage method for Spent Nuclear Fuels (SNFs), the situation monitoring technology for a DSS should be further improved to ensure the reliability of DSS during storing time. However, a passive cooling mechanism with a full-sealed storage requirement causes that the internal situation cannot be directly monitored by thermocouples inserted into the DSS. In this study, a new surveillance method, therefore, has been proposed to overcome this problem. It can predict the DSS situation through measuring the temperature profile at the Transportable Storage Canister (TSC). A validated CFD methodology has been utilized to confirm the method through simulating the thermal characteristics of the ChinShan DSS (CSDSS). The major factors, such as the thermal loading, accident situation and flaw caused by penetrated hole probably, have been considered in this present work. The result shows that the above-mentioned issues would obviously affect the temperature profile on the TSC and can be identified via detecting the temperature profile difference on of TSC. These results confirm that the indirectly surveillance method has enough capability to replace the original monitored method and provide more system information of DSS vendor for middle-term storage.

  15. 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.

  16. 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.

  17. A cask fleet operations study

    Energy Technology Data Exchange (ETDEWEB)

    1988-03-01

    This document describes the cask fleet currently available to transport spent nuclear fuels. The report describes the proposed operational procedures for these casks and the vehicles intended to transport them. Included are techniques for loading the cask, lifting it onto the transport vehicle, preparing the invoices, and unloading the cask at the destination. The document concludes with a discussion on the maintenance and repair of the casks. (tem) 29 figs.

  18. Extended Dry Storage Signature Bench Scale Detector Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-02

    This report is the conceptual design of a detector based on research within the Extended Dry Storage Signature Development project under the DOE-­NE MPACT campaign. This is the second year of the project; from this year’s positive results, the next step is building a prototype and testing with real materials .

  19. Dual purpose wheat production with different levels of nitrogen topdressing

    Directory of Open Access Journals (Sweden)

    Éderson Luis Henz

    2016-04-01

    Full Text Available Currently, the practice of Crop-Livestock Integration is stimulated as a way of increasing the generation of foreign exchange for Brazil. Integrated systems improve land use efficiency as well as preserve, recover and increment or soil fertility. The aim of this research was to evaluate how different doses of nitrogen fertilization can affect production and quality of dual purpose wheat submitted to grazing. The experimental designed was randomized block with five treatments (0, 75, 150, 225 and 300 Kg N ha-1, like ammonium nitrate and four repetitions. The forage yield, the percentage crude protein (P=.0001 and acid detergent insoluble protein (P=.0054 had a linear increased because of the nitrogen addition doses. The crude protein percentage changed the estimate of all soluble carbohydrates (P=.0001 and non-fibrous carbohydrates (P=.0186, but did not influence the, nitrogen detergent fiber corrected with ash and proteins percentage contributing for content cell. The crops production (P=.0001 and the number of kernels per ear (P=.0001 showed significantly difference because of the nitrogen additions dose, increasing the number of fertile flowers. The nitrogen topdressing alters forage production, the chemical composition and the production of dual purpose wheat grains subjected to grazing.

  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. 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.

  2. 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

  3. 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.

  4. NATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS IN A CANISTER WITH HORIZONTAL INSTALLATION OF DUAL PURPOSE CASK FOR SPENT NUCLEAR FUEL

    Directory of Open Access Journals (Sweden)

    DONG-GYU LEE

    2013-12-01

    The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to 1.5×106 1.0×107.

  5. Production indices for dual purpose cattle in central Brazil

    Directory of Open Access Journals (Sweden)

    Concepta McManus

    2011-07-01

    Full Text Available This study examined the effects of crossbreeding low genetic potential cows of Bos indicus origin characterized by Gyr crossed with Holstein-Friesian and Simmental bulls to produce animals in a low input dual purpose system. The farm is situated near Brasilia, in the savannah region of Brazil. The climate of the region is classified as Aw by Köppen. Data was available on 1580 calvings and completed lactations of cows with three genetic types: Gyr, Holstein-Friesian × Gyr and Simmental × Gyr. The bulls ran with the cows all year round and the diet comprised of pasture (mainly Brachiaria and Andropogon during the summer (rainy season and milled sugar cane with added urea during the winter (dry season. A mineral salt mixture was available ad libitum. Data was analysed using Statistical Analysis System. The results show that, under low input management conditions, the crossbred cows produce approximately twice the volume of milk per lactation, calve at a younger age and have a shorter open period, but there are no significant differences between crosses for growth rates of the calves or body condition of the cows. In this system, crossbred cows had production higher indices than zebu cattle. The best indices were found for cows calving in the rainy season (September to December and thinner cows (with body condition 3-5 on a scale of 9.

  6. System-Level Logistics for Dual Purpose Canister Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena A.

    2014-06-03

    The analysis presented in this report investigated how the direct disposal of dual purpose canisters (DPCs) may be affected by the use of standard transportation aging and disposal canisters (STADs), early or late start of the repository, and the repository emplacement thermal power limits. The impacts were evaluated with regard to the availability of the DPCs for emplacement, achievable repository acceptance rates, additional storage required at an interim storage facility (ISF) and additional emplacement time compared to the corresponding repackaging scenarios, and fuel age at emplacement. The result of this analysis demonstrated that the biggest difference in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario is for a repository start date of 2036 with a 6 kW thermal power limit. The differences are also seen in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario for the alternative with a 6 kW thermal limit and a 2048 start date, and for the alternatives with a 10 kW thermal limit and 2036 and 2048 start dates. The alternatives with disposal of UNF in both DPCs and STADs did not require additional storage, regardless of the repository acceptance rate, as compared to the reference repackaging case. In comparison to the reference repackaging case, alternatives with the 18 kW emplacement thermal limit required little to no additional emplacement time, regardless of the repository start time, the fuel loading scenario, or the repository acceptance rate. Alternatives with the 10 kW emplacement thermal limit and the DPCs and STADs fuel loading scenario required some additional emplacement time. The most significant decrease in additional emplacement time occurred in the alternative with the 6 kW thermal limit and the 2036 repository starting date. The average fuel age at emplacement ranges from 46 to 88 years. The maximum fuel age at

  7. 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.

  8. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy Of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

  9. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B. Jr.

    1999-10-21

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed.

  10. 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.

  11. 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.

  12. Source book for planning nuclear dual-purpose electric/distillation desalination plants

    Energy Technology Data Exchange (ETDEWEB)

    Reed, S.A.

    1981-02-01

    A source book on nuclear dual-purpose electric/distillation desalination plants was prepared to assist government and other planners in preparing broad evaluations of proposed applications of dual-purpose plants. The document is divided into five major sections. Section 1 presents general discussions relating to the benefits of dual-purpose plants, and spectrum for water-to-power ratios. Section 2 presents information on commercial nuclear plants manufactured by US manufacturers. Section 3 gives information on distillation desalting processes and equipment. Section 4 presents a discussion on feedwater pretreatment and scale control. Section 5 deals with methods for coupling the distillation and electrical generating plants to operate in the dual mode.

  13. 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

    different diameters and lengths would likely be on the same order of magnitude as the Basin Modifications project. The cost of a DTS capability is affected by the number of design variations of different vendor transport and dry transfer casks to be considered for design input. Some costs would be incurred for each vendor DTS to be handled. For example, separate analyses would be needed for each dry transfer cask type such as criticality, shielding, dropping a dry transfer cask and basket, handling and auxiliary equipment, procedures, operator training, readiness assessments, and operational readiness reviews. A DTS handling capability in L-Area could serve as a backup to the Shielded Transfer System (STS) for unloading long casks and could support potential future missions such as the Idaho National Laboratory (INL) Exchange or transferring UNF from wet to dry storage.

  14. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  15. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  16. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  17. 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)

  18. Comparing hemp (Cannabis sativa L.) cultivars for dual-purpose production under contrasting environments

    NARCIS (Netherlands)

    Tang, Kailei; Struik, P.C.; Yin, X.; Thouminot, C.; Bjelková, M.; Stramkale, V.; Amaducci, S.

    2016-01-01

    Interest in hemp as a multi-purpose crop is growing worldwide and for the first time in 2015 it was cultivated in Europe on more than 20.000 ha as a dual-purpose crop, for the seeds and for the fibre. In the present study, fibre and seed productivity of 14 commercial cultivars were tested in four

  19. A Community Facilities Center with Fallout Shelter as Dual Purpose Space.

    Science.gov (United States)

    Office of Civil Defense (DOD), Washington, DC.

    A presentation is made of five award-winning designs for a fireproof community recreation facility, on a selected site in New York City, incorporating a fallout shelter as a dual-purpose space. Graphic illustrations are given of the award winning designs, each of which used one of the following solutions--(1) the fallout structure above grade with…

  20. A study of thermal, structural and shielding safety analysis for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, S. H. [Kyungpook Nationl Univ., Daegu (Korea, Republic of)

    1997-03-15

    As a replaced method for MRS, the dry storage has been intensively developed by the advanced countries of nuclear power technology. Currently, the domestic technology for the dry storage is also under development. In the present study, the developed technical standards for USNRC and its operation are summarized. Futhermore, the SAR for VECTRA's NUHOMES satisfied with DOE and NRC's requirements is inversely analyzed and combined with both USNRC's regulatory guide and LLNL's SARS. In the safety analysis of a dry storage, the principal design criteria which identifies the structural and mechanical safety criteria is investigated. Based on the design criteria, hypothetical accident analysis as well as off-normal operation analysis are investigated.

  1. Design requirements of a consolidating dry storage module for CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Yang, Ke Hyung; Choi, Byung Il; Lee, Heung Young [KHNP/NETEC, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-10-01

    This paper presents a technical description of design requirement document covers the requirements of the MACSTOR/KN-400 module, which is under development to densely accommodate CANDU spent fuels with more efficient way. The design requirement is for the module that will be constructed within a dry storage site after successfully licensed by the regulatory body. This temporary outdoor spent fuel dry storage facility provides for safe storage of spent nuclear fuel after it has been removed from the plant's storage pool after being allowed to decay for a period of at least 6 years. The MACSTOR/KN-400 module is being designed to the envelope of site environmental conditions encountered at the Wolsong station. The design requirements of MACSTOR/KN-400 module meets the requirements of the appropriate Codes and Standards for dry storage of spent fuel from nuclear power reactors such as lOCFR72, and Korea Atomic Energy Act and relevant technical standard.

  2. 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.

  3. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling).

  4. Critical review of creep FRAPCON-3 model under dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E. [Unit of Nuclear Safety Research, CIEMAT, Avda. Complutense 22, Madrid, Madrid 28040 (Spain)

    2009-06-15

    There is a general agreement that cladding creep rupture is the most likely and limiting failure mechanism of spent fuel in dry storage compared to other potential mechanisms, like stress corrosion cracking and/or delayed hydride cracking. Nevertheless, occurrence of creep rupture is very improbable since both decay heat and hoop stress tend to decrease throughout dry storage. In spite of this, the current trend to higher burn up levels needs further attention that ensures safe storage of spent fuel irradiated over 45 GWd/MTU. An extensive work has been carried out during the last four decades in the area of in-reactor creep modelling. Unfortunately, the in-reactor conditions are so different from those prevailing under dry storage, that all the experience gained cannot be extrapolated in a straightforward manner. On the other side, as creep tests simulating conditions throughout a 20-40 year dry storage are impractical, post-irradiation cladding creep behaviour has been modelled by means of time-temperature dependent laws developed on the basis of currently available zirconium alloys data. Additionally, some tests have been exploring the effect of irradiation, hydrogen distribution and material composition on the materials creep behaviour. Adaptation of fuel performance codes initially developed for normal and off-normal reactor operation is not an easy task either. Creep modelling is usually dependent of host codes because a good part of its validation and update has been carried out in an integral way, and as a consequence its independent performance assessment is not an easy task. This work examines the current capability of FRAPCON-3 to model creep behaviour under dry storage conditions. To do so, a review of its major fundamentals has been done and its range of applicability discussed. Once its main approximations and drawbacks have been identified, an attempt to overcome some of them has been intended by implementing an alternative expression for creep under

  5. Genetics Home Reference: CASK-related intellectual disability

    Science.gov (United States)

    ... Conditions CASK-related intellectual disability CASK-related intellectual disability Printable PDF Open All Close All Enable Javascript ... the expand/collapse boxes. Description CASK -related intellectual disability is a disorder of brain development that has ...

  6. 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.

  7. 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.

  8. DESIGN OF A COMPACT DUAL-PURPOSE STARTING CLUTCH IN THE DRIVE OF A PROTOTYPE VEHICLE

    Directory of Open Access Journals (Sweden)

    Ján PETRÓCI

    2016-06-01

    Full Text Available Initially, the development of a dual-purpose clutch was based on racing experiences and application requirements, as well as the results from testing the new power unit in the existing prototype vehicle. In order to achieve the highest possible driving range of the prototype vehicle, it has been necessary to eliminate the maximum possible losses and drive in unnecessary components. The design aimed to achieve simple access, reliability and low weight.

  9. Dual-purpose RTOF diffractometer facility at the ET-RR-1 reactor. Research note

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R.M.A.; Hiismaeki, P.E.; Antson, O.K.; Tiitta, A.T.

    1992-01-01

    Some new features of an RTOF (Reverse Time of Flight) diffractometer facility, to be installed at the ET-RR-1 reactor, are considered. The suggested facility will be a dual-purpose instrument. It can be used for high-resolution crystallography using transmission diffraction technique and for strain measurements of technical components using a scattering geometry with a detector at 90 deg. (Copyright (c) Valtion teknillinen tutkimuskeskus (VTT) 1992.)

  10. Dual-purpose RTOF diffractometer facility at the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R.M.A. [Atomic Energy Authority, Cairo (Egypt); Hiisimaeki, P.E.; Antson, O.K.; Tiitta, A.T. [Technical Research Centre of Finland, Espoo (Finland). Reactor Lab.

    1992-11-01

    Some new features of an RTOF diffractometer facility, to be installed at the ET-RR-1 reactor, are considered. The suggested facility will be a dual-purpose instrument. It can be used for high-resolution crystallography utilizing transmission diffraction technique and for strain measurements of technical components using a scattering geometry with a detector at 90 deg. (orig.). (30 refs., 1 fig.).

  11. Evaluation of limiting mechanisms for long-term spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J. [ANATECH Research Corp., San Diego, CA (United States); Machiels, A. [EPRI, Palo Alto, CA (United States)

    2001-07-01

    Several failure mechanisms have been postulated that could become limiting for spent fuel in dry storage. These are: stress Corrosion Cracking (SCC), Delayed Hydride Cracking (DHC) and Creep Rupture (CR). These mechanisms are examined in some detail from two perspectives: their initial environments in which they were developed and applied, and in relation to their applicability to dry storage. Extrapolation techniques are used to transfer the mechanisms from their initial in-reactor and laboratory domains to out-of-reactor spent fuel dry storage environments. This transfer is accomplished both qualitatively where necessary and quantitatively when possible, with fracture toughness used as the transfer function. In this regard, the paper provides useful information on cladding fracture toughness estimates that recognize the specific physical conditions of the cladding, which would not be found elsewhere in the literature. The arguments presented in this paper confirm the general technical consensus that creep is the governing mechanism for spent fuel in long-term dry storage. (author)

  12. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  13. A study on safety analysis methodology in spent fuel dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

    2004-02-15

    Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

  14. 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.

  15. Shielding Analysis of the 5320 Shipping Cask

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A. [Westinghouse Savannah River Company, AIKEN, SC (United States); Nathan, S. [Westinghouse Safety Management Solutions, Aiken, SC (United States)

    1998-05-01

    The purpose of this work is to demonstrate that the 5320 shipping cask meets Federal regulations for maximum radiation dose rates when loaded with the intended plutonium oxide cargo. It should be emphasized that the 5320 is an existing cask, and therefore this work represents confirmatory analysis rather than design analysis.

  16. CFD Analysis of a Dry Storage System for MACSTOR/KN-400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yu sun; Shin, Byung soo; Chang, Soon heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2007-10-15

    There are four nuclear power plants in operation at Wolsong and, annually, more than 5,000 assemblies of spent nuclear fuels are released from each plant. Thus the concrete silo type dry storage system was constructed from '90. However, after 2006, another dry storage facility is required to accept the all amount of spent nuclear fuels from the plants. Instead of the concrete silo type, MACSTOR/KN-400 was developed to store the CANDU spent nuclear fuel more densely. In this study, computational fluid dynamics (CFD) analysis of the MACSTOR/KN-400 model was performed to confirm the thermal integrity of the facility, especially for the concrete structure, using the commercial CFD code, FLUENT. The MACSTOR/KN-400 which has Thermal Insulation Panel (TIP) and IAEA Re-verification Pipe (RVP) was modeled and analyzed in the view point of the thermal integrity of the concrete structure.

  17. Seismic analysis and design of a spent nuclear fuel dry storage system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kyu-Sup; Jeong, In-Su; Kim, Kap-Sun; Kim, Jong-Soo [KONES Corporation, Seoul (Korea, Republic of); Yoon, Jeong-Hyoun; Kang, Young-Gon; Cho, Chun-Hyung; Lee, Heung-Young [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-11-15

    A comprehensive seismic analysis has been conducted in this paper by using the computer programs, SHAKE and SASSI, for a consolidated dry storage module named MACSTOR/KN-400, which will be constructed in Wolsung CANDU Nuclear Power Plant site. Especially seismic soil-structure interaction effects have been investigated in terms of transfer functions, maximum floor acceleration and floor response spectrum, and finally those effects are to be incorporated in the detailed structure design.

  18. Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

    1998-07-01

    This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

  19. 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.

  20. Select Generic Dry-Storage Pilot Plant Design for Safeguards and Security by Design (SSBD) per Used Fuel Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Demuth, Scott Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sprinkle, James K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-26

    As preparation to the year-end deliverable (Provide SSBD Best Practices for Generic Dry-Storage Pilot Scale Plant) for the Work Package (FT-15LA040501–Safeguards and Security by Design for Extended Dry Storage), the initial step was to select a generic dry-storage pilot plant design for SSBD. To be consistent with other DOE-NE Fuel Cycle Research and Development (FCR&D) activities, the Used Fuel Campaign was engaged for the selection of a design for this deliverable. For the work Package FT-15LA040501–“Safeguards and Security by Design for Extended Dry Storage”, SSBD will be initiated for the Generic Dry-Storage Pilot Scale Plant described by the layout of Reference 2. SSBD will consider aspects of the design that are impacted by domestic material control and accounting (MC&A), domestic security, and international safeguards.

  1. Evaluation of a Dual-Purpose Chemical Applicator for Paddy Fields

    Directory of Open Access Journals (Sweden)

    Mohammed S. Abubakar

    2011-01-01

    Full Text Available Problem statement: One of the major problems of rice production is the shortage of labor resulting from migration from rural to urban areas, making it very difficult to meet peak demands for paddy production. In most developing countries of Southeast Asia, agricultural mechanization of paddy field operation is mostly carried out using conventional machines. For example in rice production powerintensive operations such as water pumping, land preparations, transplanting seedlings, harvestings and threshing are being mechanized but other operations like fertilizer and chemical (pesticides applications are not yet fully mechanized, rather they are performed manually with motorized backpack knapsack sprayers which have many disadvantages. The main objective of this study was to develop and evaluate a new concept dual-purpose chemical applicator for paddy fields farmers in order to complement the labor shortage during the peak period. Approach: A dual-purpose chemical applicator for flooded paddy field was evaluated using the S341.4 Standard in respect to the distribution patterns/droplet sizes and uniformity of spreading/spraying for the chemical application to boost agricultural mechanization in rice production and also to overcome the safety concern of hazardous spray drift during chemical application by the paddy farmers. The dual purpose applicator was mounted on a high clearance prime mover. The machine performances for both granular and liquid chemical application were reported. Results: Results for urea granular chemical indicate that at low (40 kg ha−1 and high (120 kg ha−1 rates and 550 rpm disc speed distribution pattern skewed to the left whereas the distribution pattern shape at medium (80 kg ha−1 rates was good flattop. At high rate (120 kg ha−1 and 1000 rpm disc speed, mean distribution pattern became poor (M-shape and also at low and medium application rates the distribution pattern shapes lopsided

  2. Crossbreeding dual-purpose cattle for beef production in tropical regions.

    Science.gov (United States)

    Rodriguez-Voigt, A; Noguera, E; Rodriguez, H L; Huerta-Leidenz, N O; Morón-Fuenmayor, O; Rincón-Urdaneta, E

    1997-11-01

    Six breed types groups of six steer calves each were used to observe differences in growth and carcass traits: F1 Brown Swiss (F1BS), F1 Holstein (F1HO), F2 Brahman (F2BR), F2 Criollo Rio Limon (F2CRL), purebred Criollo Rio Limon (CRL), and Perija Mosaic (PMO). After 404 days of grazing, the heaviest F1BS and F2BR steers were different from F1HO and CRL counterparts (p carcass maturity, quality or yield grade, boneless cut percentages, bone percentage or percent trimmable fat did not vary among breed types. All steers were within the A maturity level and graded Standard. Results indicate the importance of feeding dual-purpose steers to heavier weights to please industry preferences.

  3. Limitations and potentials of dual-purpose cow herds in Central Coastal Veracruz, Mexico.

    Science.gov (United States)

    Absalón-Medina, Victor Antonio; Blake, Robert W; Fox, Danny Gene; Juárez-Lagunes, Francisco I; Nicholson, Charles F; Canudas-Lara, Eduardo G; Rueda-Maldonado, Bertha L

    2012-08-01

    Feed chemical and kinetic composition and animal performance information was used to evaluate productivity limitations and potentials of dual-purpose member herds of the Genesis farmer organization of central coastal Veracruz, Mexico. The Cornell Net Carbohydrate and Protein System model (Version 6.0) was systematically applied to specific groups of cows in structured simulations to establish probable input-output relationships for typical management, and to estimate probable outcomes from alternative management based on forage-based dietary improvements. Key herd vulnerabilities were pinpointed: chronic energy deficits among dry cows of all ages in late gestation and impeded growth for immature cows. Regardless of the forage season of calving, most cows, if not all, incur energy deficits in the final trimester of gestation; thus reducing the pool of tissue energy and constraining milking performance. Under typical management, cows are smaller and underweight for their age, which limits feed intake capacity, milk production and the probability of early postpartum return to ovarian cyclicity. The substitution of good-quality harvested forage for grazing increased predicted yields by about one-third over typical scenarios for underweight cows. When diets from first parturition properly supported growth and tissue repletion, milk production in second and third lactations was predicted to improve about 60%. Judiciously supplemented diets based on good quality grass and legume forages from first calving were predicted to further increase productivity by about 80% across a three-lactation cow lifetime. These dual-purpose herd owners have large incentives to increase sales income by implementing nutritional strategies like those considered in this study.

  4. 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)

  5. The Analytical Review of the Condition of Heavy Class Military and Dual-Purpose Unmanned Ground Vehicle

    Directory of Open Access Journals (Sweden)

    Blokhin Aleksandr

    2015-01-01

    Full Text Available The purpose of this article is the evaluation of the actual condition of heavy (weight more than 700 kg military robotics and dual-purpose robotics in the world. The extensive review of the world market of heavy class military unmanned ground vehicle was made. All reviewed robots are used at present time or exist like prototypes. All robots were systematized by most important technical characteristics. In the closing of article the analysis of the reviewed heavy class dual purpose UGVs are presented. Based on the analysis the conclusion about actual condition of the heavy military robotics and dual-purpose robotics was made. Also the most promising ways and tendencies of development are representeds.

  6. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  7. A commentary on domestic animals as dual-purpose models that benefit agricultural and biomedical research.

    Science.gov (United States)

    Ireland, J J; Roberts, R M; Palmer, G H; Bauman, D E; Bazer, F W

    2008-10-01

    Research on domestic animals (cattle, swine, sheep, goats, poultry, horses, and aquatic species) at land grant institutions is integral to improving the global competitiveness of US animal agriculture and to resolving complex animal and human diseases. However, dwindling federal and state budgets, years of stagnant funding from USDA for the Competitive State Research, Education, and Extension Service National Research Initiative (CSREES-NRI) Competitive Grants Program, significant reductions in farm animal species and in numbers at land grant institutions, and declining enrollment for graduate studies in animal science are diminishing the resources necessary to conduct research on domestic species. Consequently, recruitment of scientists who use such models to conduct research relevant to animal agriculture and biomedicine at land grant institutions is in jeopardy. Concerned stakeholders have addressed this critical problem by conducting workshops, holding a series of meetings with USDA and National Institutes of Health (NIH) officials, and developing a white paper to propose solutions to obstacles impeding the use of domestic species as dual-purpose animal models for high-priority problems common to agriculture and biomedicine. In addition to shortfalls in research support and human resources, overwhelming use of mouse models in biomedicine, lack of advocacy from university administrators, long-standing cultural barriers between agriculture and human medicine, inadequate grantsmanship by animal scientists, and a scarcity of key reagents and resources are major roadblocks to progress. Solutions will require a large financial enhancement of USDA's Competitive Grants Program, educational programs geared toward explaining how research using agricultural animals benefits both animal agriculture and human health, and the development of a new mind-set in land grant institutions that fosters greater cooperation among basic and applied researchers. Recruitment of

  8. Economic evaluation of dual purpose desalination plants by fuel type in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Seung-Su, Kim; Man-Ki, Lee [Korea Atomic Energy Research Institute, Dae-jeon city (Korea, Republic of)

    2007-07-01

    In light of the recent rapid increase in the fossil fuel prices it is meaningful to evaluate the impact of these price changes in the economics of dual-purpose desalination projects producing electricity and fresh water simultaneously. The price of crude oil and LNG (Liquefied Natural Gas) has increased by about 200% and 100% during the past three or four years. The uranium price has also increased by nearly 500% during the same period. The purpose of this paper is to analyze and compare the economics of SMART (System-integrated Modular Advanced ReacTor) which is being developed as a small size PWR type and the LNG Combine Cycle coupled with MED (Multi-Effect Distillation) which are being acknowledged as promising energy sources for the future in Korea. The methods of analysis used in this paper are the lifetime leveled cost method for the power and water cost calculation and the power credit method for the total cost allocation. DEEP (Devaluation Economic Evaluation Program) developed by IAEA was used to perform an economic comparison between the two dual-purpose desalination projects. From the results of the analysis it is found that the desalination by SMART-MED is much superior to that of LNG CC-MED under the current economic and technical situations. It is also shown that the relative superiority of SMART-MED to LNG CC-MED will be maintained even in case where an increase in the uranium price and the SMART construction cost would reach a maximum in the sensitivity analysis. In the case that the discount rate declines to 5% per year, the relative attractiveness of SMART-MED which is a capital intensive plant will be enhanced when compared to that for a 7% discount rate. In addition to this, it is thought that a nuclear energy source will be favored much more than now in the field of desalination if the regulations for the emission of greenhouse gases are to be strengthened. (authors)

  9. 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.

  10. Feed and fuel: the dual-purpose advantage of an industrial sweetpotato.

    Science.gov (United States)

    Mussoline, Wendy A; Wilkie, Ann C

    2017-03-01

    Sustainable agricultural systems must support nutritional requirements, meet the energy demands of a growing population, preserve environmental resources and mitigate climate change. The sweetpotato (Ipomoea batatas L.) is a high-yielding crop that requires minimal fertilization and irrigation, and the CX-1 industrial cultivar offers superior potential for feed and fuel. CX-1 had the highest agronomic fresh vine yield (51.5 t ha(-1) ), averaged over two cropping seasons, compared with Hernandez (33.7) and Beauregard (21.8) varieties. CX-1 vines were more nutritional than the table varieties, specifically in regard to relative feed value (205), water-soluble carbohydrates (171 g kg(-1) dry matter (DM)), total digestible nutrients (643 g kg(-1) DM), metabolizable energy (10.2 MJ kg(-1) DM) and organic matter digestibility. Their lower fiber and lignin concentrations contributed to their freshness and digestibility throughout maturity. Significantly higher iron concentrations make the CX-1 vines a valuable, low-fat iron supplement for animal feed. The CX-1 roots also showed the highest bioethanol potential (82.3 g ethanol kg(-1) fresh root) compared to Hernandez (64.5) and Beauregard (48.1). The CX-1 industrial sweetpotato is an ideal dual-purpose crop for tropical/subtropical climates that can be utilized as a non-grain-based feedstock for bioethanol production while contributing a valuable, high-yielding nutritional supplement for animal feed. © 2016 Society of Chemical Industry. © 2016 Society of Chemical Industry.

  11. Genetic relationships between dagginess, breech bareness, and wool traits in New Zealand dual-purpose sheep.

    Science.gov (United States)

    Pickering, N K; Blair, H T; Hickson, R E; Dodds, K G; Johnson, P L; McEwan, J C

    2013-10-01

    Genetic and phenotypic parameters were estimated for dagginess, breech, wool, and fiber traits from approximately 29,500 progeny born in 2009 and 2010 in New Zealand dual-purpose ram breeding sheep flocks. Dagginess is adherence of fecal matter to the wool, and this study investigates the genetic and phenotypic correlations between dagginess and breech and wool traits. Estimates for heritability were moderate (0.21 to 0.44) for the following traits: dag score at 3 and 8 mo (DAG3, DAG8), breech bareness, wool length, wool bulk (BULK), mean fiber diameter, mean fiber diameter SD, mean fiber diameter CV, curvature (CURV), weaning weight at 3 mo, and autumn BW. Heritability estimates for fleece weight at 12 mo and proportion of medullated fibers were high (0.49 and 0.53, respectively). Dag score at 3 mo and DAG8 had low genetic and phenotypic correlations with all traits. Breech bareness had positive genetic and phenotypic correlations with CURV and BULK and mostly negative genetic correlations with all other wool traits. In summary the quantity and attributes of wool were not primary causative factors in fecal accumulation, leaving fecal consistency and composition as the major factors.

  12. Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

    Energy Technology Data Exchange (ETDEWEB)

    Trond Bjornard; Philip C. Durst

    2012-05-01

    This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA) of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement with

  13. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  14. 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.

  15. Structural design concept and static analysis of CANDU spent fuel compact dry storage system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y. [Korea Hydro and Nuclear Power Company, Taejon (Korea, Republic of)

    2003-07-01

    In this study, an structural design concept on CANDU spent fuel compact dry storage system MACSTOR/KN-400 module has been established with a view to optimally design the structural members of the system. Design loads, loading combination and structural safety criteria of the module were reviewed assuming W olsung Site. The static analysis of the module showed that compressive stress concentration due to dead load and live load occurred around the center of roof slab. Maximum stress resulted from dead load is about twice as much as the stress from live load, and structural behavior of module caused by wind load was not significant. The static analysis results will have influence on the reinforcement bar design of structural members with other structural analyses.

  16. ACR fuel storage analysis: finite element heat transfer analysis of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Khair, K.; Baset, S.; Millard, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Over the past decade Atomic Energy of Canada Limited (AECL) has designed and licensed air-cooled concrete structures used as above ground dry storage containers (MACSTOR) to store irradiated nuclear fuel from CANDU plants. A typical MACSTOR 200 module is designed to store 12,000 bundles in 20 storage cylinders. MACSTOR 200 modules are in operation at Gentilly-2 in Canada and at Cernavoda in Romania. The MACSTOR module is cooled passively by natural convection and by conduction through the concrete walls and roof. Currently AECL is designing the Advanced Candu Reactor (ACR) with CANFLEX slightly enriched uranium fuel to be used. AECL has initiated a study to explore the possibility of storing the irradiated nuclear fuel from ACR in MACSTOR modules. This included work to consider ways of minimizing footprint both in the spent fuel storage bay and in the dry storage area. The commercial finite element code ANSYS has been used in this study. The FE model is used to complete simulations with the higher heat source using the same concrete structural dimensions to assess the feasibility of using the MACSTOR design for storing the ACR irradiated fuel. This paper presents the results of the analysis. The results are used to confirm the possibility of using, with minimal changes to the design of the storage baskets and the structure, the proven design of the MACSTOR 200 containment to store the ACR fuel bundles with higher enrichment and burnup. This has thus allowed us to confirm conceptual feasibility and move on to investigation of optimization. (author)

  17. Seismic Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Cho, Chun Hyung; Lee, Heung Young [Korea Hydro and Nuclear Power Co., Ltd., Taejon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su; Kim, Jong Soo [KONES Co., Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside of the concrete module are built 40 storage cylinders accommodating ten 60- bundle dry storage baskets, which are suspended from the top slab and eventually constrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module is by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants except for local geologic characteristics. As per USNRC SRP Section 3.7.2 and current US practices, Soil-Structure Interaction (SSI) effect shall be considered for all structures not supported by a rock or rock-like soil foundation materials. An SSI is a very complicated phenomenon of the structure coupled with the soil medium that is usually semi-infinite in extent and highly nonlinear in its behavior. And the effect of the SSI is noticeable especially for stiff and massive structures resting on relatively soft ground. Thus the SSI effect has to be considered in the seismic design of MACSTOR/KN-400 module resting on soil medium. The scope of the this paper is to carry out a seismic SSI analysis of the MACSTOR/KN-400 module, in order to show how much the SSI gives an effect on the structural responses by comparing with the fixed-base analysis.

  18. Evaluation of tropical grasses for milk production by dual-purpose cows in tropical Mexico.

    Science.gov (United States)

    Juarez Lagunes, F I; Fox, D G; Blake, R W; Pell, A N

    1999-10-01

    Two experiments using the Cornell Net Carbohydrate and Protein System were conducted to characterize the carbohydrate and protein fractions and corresponding rates of digestion of 15 tropical pasture grasses and to evaluate their ability to support milk production by dual-purpose cows. In the first experiment, ranges in carbohydrate and protein fractions of 15 grasses at 35 to 42 d of regrowth were: neutral detergent fiber (NDF) 63.5 to 74.9% of DM; permanganate lignin 4.7 to 7.8% of NDF; CP 5.5 to 11.9% of DM; and soluble protein 15.1 to 44.1% of crude protein (CP). The ranges of rates of digestion expressed as percent per hour were neutral detergent solubles (7.5 to 27.4); NDF (3.8 to 8.4); and neutral detergent insoluble protein (2.9 to 9.5). Predictions of the amount of milk that could be produced based on the amount of metabolizable energy supplied by the diet decreased 35% when NDF increased from 60 to 80%, and increased 88% when the rate of digestion of NDF increased from 3 to 6%/h. The milk production that could be sustained based on metabolizable protein in the diet doubled as CP increased from 4 to 12%. In the second experiment, nitrogen fertilization reduced NDF 7.3% and increased CP 84% without changing protein solubility, resulting in increased rumen nitrogen and metabolizable protein balances. With all forages, the Cornell Net Carbohydrate and Protein System predicted that availability of metabolizable protein would limit milk production. Predicted microbial growth was limited by ruminally available protein rather than by available carbohydrate.

  19. 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.

  20. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  1. A model on the effect of temperature and moisture on pollen longevity in air-dry storage environments

    NARCIS (Netherlands)

    Hong, T.D.; Ellis, R.H.; Buitink, J.; Walters, J.; Hoekstra, F.A.; Crane, J.

    1999-01-01

    Data on the survival of pollen ofTypha latifoliaL. stored for up to 261 d over seven different saturated salt solutions (providing 0.5 to 66% relative humidity) and six different constant temperatures (from −5 to +45 °C) were analysed to quantify the effect of air-dry storage environment on pollen l

  2. Economic values for health and feed efficiency traits of dual-purpose cattle in marginal areas.

    Science.gov (United States)

    Krupová, Z; Krupa, E; Michaličková, M; Wolfová, M; Kasarda, R

    2016-01-01

    Economic values of clinical mastitis, claw disease, and feed efficiency traits along with 16 additional production and functional traits were estimated for the dairy population of the Slovak Pinzgau breed using a bioeconomic approach. In the cow-calf population (suckler cow population) of the same breed, the economic values of feed efficiency traits along with 15 further production and functional traits were calculated. The marginal economic values of clinical mastitis and claw disease incidence in the dairy system were -€ 70.65 and -€ 26.73 per case per cow and year, respectively. The marginal economic values for residual feed intake were -€ 55.15 and -€ 54.64/kg of dry matter per day for cows and breeding heifers in the dairy system and -€ 20.45, -€ 11.30, and -€ 6.04/kg of dry matter per day for cows, breeding heifers, and fattened animals in the cow-calf system, respectively, all expressed per cow and year. The sums of the relative economic values for the 2 new health traits in the dairy system and for residual feed intake across all cattle categories in both systems were 1.4 and 8%, respectively. Within the dairy production system, the highest relative economic values were for milk yield (20%), daily gain of calves (20%), productive lifetime (10%), and cow conception rate (8%). In the cow-calf system, the most important traits were weight gain of calves from 120 to 210 d and from birth to 120 d (19 and 14%, respectively), productive lifetime (17%), and cow conception rate (13%). Based on the calculation of economic values for traits in the dual-purpose Pinzgau breed, milk production and growth traits remain highly important in the breeding goal, but their relative importance should be adapted to new production and economic conditions. The economic importance of functional traits (especially of cow productive lifetime and fertility) was sufficiently high to make the inclusion of these traits into the breeding goal necessary. An increased interest

  3. Estimation and comparison of benefits due to feeding hay and silage during the dry season on commercial dual-purpose cattle production systems in Honduras and Costa Rica

    NARCIS (Netherlands)

    Schoonhoven, A.D.; Holmann, F.; Argel, P.; Ordoñez, J.C.; Chaves, J.

    2006-01-01

    Smallholders with dual-purpose cattle production systems in most Central America experience a shortage of forages during the dry season (4-8 month. As a result, substantially lower milk production and weight gain occurs. Dual-purpose operations seeking to maximize milk and beef production in the dry

  4. Assessment of xylanase activity in dry storage as a potential method of reducing feedstock cost.

    Science.gov (United States)

    Smith, William A; Thompson, David N; Thompson, Vicki S; Radtke, Corey W; Carter, Brady

    2009-05-01

    Enzymatic preprocessing of lignocellulosic biomass in dry storage systems has the potential to improve feedstock characteristics and lower ethanol production costs. To assess the potential for endoxylanase activity at low water contents, endoxylanase activity was tested using a refined wheat arabinoxylan substrate and three commercial endoxylanases over the water activity range 0.21-1.0, corresponding to water contents of 5% to >60% (dry basis). Homogeneously mixed dry samples were prepared at a fixed enzyme to substrate ratio and incubated in chambers at a variety of fixed water activities. Replicates were sacrificed periodically, and endoxylanase activity was quantified as an increase in reducing sugar relative to desiccant-stored controls. Endoxylanase activity was observed at water activities over 0.91 in all enzyme preparations in less than 4 days and at a water activity of 0.59 in less than 1 week in two preparations. Endoxylanase activity after storage was confirmed for selected desiccant-stored controls by incubation at 100% relative humidity. Water content to water activity relationships were determined for three lignocellulosic substrates, and results indicate that two endoxylanase preparations retained limited activity as low as 7% to 13% water content (dry basis), which is well within the range of water contents representative of dry biomass storage. Future work will examine the effects of endoxylanase activity toward substrates such as corn stover, wheat straw, and switchgrass in low water content environments.

  5. NDE to Manage Atmospheric SCC in Canisters for Dry Storage of Spent Fuel: An Assessment

    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); Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, Aaron A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-01

    This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

  6. Recent findings on the oxidation of UO{sub 2} fuel under nominally dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P.; McEachern, R.J.; Sunder, S.; Wasywich, K.M.; Miller, N.H.; Wood, D.D

    1995-07-01

    This paper is an overview of fuel-storage demonstration experiments, supporting research on UO{sub 2} oxidation, and associated model development, in progress at AECL's Whiteshell Laboratories. The work is being performed to determine the time/temperature limits for safe storage of irradiated CANDU fuel in dry air. The most significant recent experimental finding has been the detection of small quantities of U{sub 3}O{sub 8}, formed over periods of one to several years in a variety of experiments at 150-170 deg C. Another important trading is the slight suppression of U{sub 3}O{sub 8} formation in SIMFUEL and other doped U0{sub 2} formulations. The development of a nucleation-and-growth model for U{sub 3}O{sub 8} formation is discussed, along with available activation energy data. These provide a basis for predicting U{sub 3}O{sub 8} formation rates under dry-storage conditions, and hence optimizing fuel storage strategies. (author)

  7. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository; Yucca Mountain Site characterization project

    Energy Technology Data Exchange (ETDEWEB)

    Hartman, D.J.; Miller, D.D. [Bechtel National, Inc., San Francisco, CA (United States); Hill, R.R. [Sandia National Labs., Albuquerque, NM (United States)

    1992-04-01

    This study investigates spent fuel cask handling experience at existing nuclear facilities to determine appropriate cask decontamination and maintenance operations at the potential Yucca Mountain repository. These operations are categorized as either routine or nonroutine. Routine cask decontamination and maintenance tasks are performed in the cask preparation area at the repository. Casks are taken offline to a separate cask maintenance area for major nonroutine tasks. The study develops conceptual designs of the cask preparation area and cask maintenance area. The functions, layouts, and major features of these areas are also described.

  8. Protective coatings for the Ontario Power Generation Inc. dry storage container

    Energy Technology Data Exchange (ETDEWEB)

    Lewak, R. [Kinectrics Inc., Toronto, Ontario (Canada)]. E-mail: richard.lewak@kinectrics.com; Ellsworth, B. [Ontario Power Generation Inc., Tiverton, Ontario (Canada)]. E-mail: brad.ellsworth@opg.com

    2006-07-01

    Ontario Power Generation (OPG), formerly Ontario Hydro, has been storing used CANDU fuel in the irradiated fuel bays (IFBs) at its nuclear generating stations since they began operation. As the IFBs began reaching capacity, the dry storage of previously cooled used fuel became an economically viable alternative to the construction of additional wet fuel bays and the OPG Dry Storage Container (DSC) was developed. The OPG DSC is a free standing reinforced concrete container, with an inner and outer carbon steel shell, for the storage and transportation of used CANDU fuel. The outer steel shell of the DSC is protected by an applied coating system to facilitate decontamination of the outer shell and to provide protective corrosion resistance. In 1990 a study was performed to determine the optimal commercial coating system to be considered as a protective coating on the DSC outer shell. An experimental program was undertaken to identify the optimal commercial coating system, which had the best decontamination characteristics as well as maximum resistance to abrasion, weathering and durability. A total of nine coating systems were selected for study, five epoxy coatings, three epoxy/polyurethane coatings, and one polyurethane coating. Sand blasted carbon steel coupons, similar to the DSC outer shell, were coated by the manufacturers and submitted for testing of the decontamination characteristics such as activity uptake, decontamination of the coating, and the activity 'sweating' phenomenon. Tests identified four commercially available pure epoxy and epoxy/polyurethane protective coating systems as being the most promising for the exterior of the DSC surface. Of these four, the coating system chosen for use on the DSC was an epoxy/polyurethane system. After a decade of use, however, several safety and environmental concerns centering on the isocyanate content present in the polyurethane and the Volatile Organic Component (VOC) content of the coating system have

  9. 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.

  10. CFD Analysis on the Passive Heat Removal by Helium and Air in the Canister of Spent Fuel Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Do Young; Jeong, Ui Ju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    In the current commercial design, the canister of the dry storage system is mainly backfilled with helium gas. Helium gas shows very conductive behavior due to high thermal conductivity and small density change with temperature. However, other gases such as air, argon, or nitrogen are expected to show effective convective behavior. Thus these are also considered as candidates for the backfill gas to provide effective coolability. In this study, to compare the dominant cooling mechanism and effectiveness of cooling between helium gas and air, a computational fluid dynamics (CFD) analysis for the canister of spent fuel dry storage system with backfill gas of helium and air is carried out. In this study, CFD simulations for the helium and air backfilled gas for dry storage system canister were carried out using ANSYS FLUENT code. For the comparison work, two backfilled fluids were modeled with same initial and boundary conditions. The observed major difference can be summarized as follows. - The simulation results showed the difference in dominant heat removal mechanism. Conduction for helium, and convection for air considering Reynolds number distribution. - The temperature gradient inside the fuel assembly showed that in case of air, more effective heat mixing occurred compared to helium.

  11. Advanced development in phytochemicals analysis of medicine and food dual purposes plants used in China (2011-2014).

    Science.gov (United States)

    Zhao, Jing; Ge, Li-Ya; Xiong, Wei; Leong, Fong; Huang, Lu-Qi; Li, Shao-Ping

    2016-01-08

    In 2011, we wrote a review for summarizing the phytochemical analysis (2006-2010) of medicine and food dual purposes plants used in China (Zhao et al., J. Chromatogr. A 1218 (2011) 7453-7475). Since then, more than 750 articles related to their phytochemical analysis have been published. Therefore, an updated review for the advanced development (2011-2014) in this topic is necessary for well understanding the quality control and health beneficial phytochemicals in these materials, as well as their research trends.

  12. 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.

  13. 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.

  14. Assessment of the sustainability of dual-purpose farms by the IDEA method in the subtropical area of central Mexico.

    Science.gov (United States)

    Salas-Reyes, Isela Guadalupe; Arriaga-Jordán, Carlos Manuel; Rebollar-Rebollar, Samuel; García-Martínez, Anastacio; Albarrán-Portillo, Benito

    2015-08-01

    The objective of this study was to assess the sustainability of 10 dual-purpose cattle farms in a subtropical area of central Mexico. The IDEA method (Indicateurs de Durabilité des Exploitations Agricoles) was applied, which includes the agroecological, socio-territorial and economic scales (scores from 0 to 100 points per scale). A sample of 47 farms from a total of 91 registered in the local livestock growers association was analysed with principal component analysis and cluster analysis. From results, 10 farms were selected for the in-depth study herein reported, being the selection criterion continuous milk production throughout the year. Farms had a score of 88 and 86 points for the agroecological scale in the rainy and dry seasons. In the socio-territorial scale, scores were 73 points for both seasons, being the component of employment and services the strongest. Scores for the economic scale were 64 and 56 points for the rainy and dry seasons, respectively, when no economic cost for family labour is charged, which decreases to 59 and 45 points when an opportunity cost for family labour is considered. Dual-purpose farms in the subtropical area of central Mexico have a medium sustainability, with the economic scale being the limiting factor, and an area of opportunity.

  15. 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...

  16. 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.

  17. A comparison between slaughter traits and meat quality of various sheep breeds: wool, dual-purpose and mutton.

    Science.gov (United States)

    Cloete, J J E; Hoffman, L C; Cloete, S W P

    2012-07-01

    The slaughter and meat quality traits of 20-month-old wool (Merino), dual-purpose (Dohne Merino and South African Mutton Merino [SAMM]) and mutton (Dormer) type sheep were compared. Average live weights of SAMM and Dormer sheep were 23% heavier than those of Dohne Merinos which were 28% heavier than Merinos. Fat depths at the thirteenth rib and lumbar regions of Merino and Dohne Merino sheep were lower than those of SAMM and Dormer sheep. The cooking loss, drip loss and shearing value from the M. longissimus dorsi did not differ between breeds. The initial juiciness and sustained juiciness of meat from Merinos were rated significantly lower by sensory analysis. Meat from Dohne Merino was rated significantly more tender for the attribute first bite. It was demonstrated that Dormer and SAMM sheep had heavier but fatter carcasses than Merinos and Dohne Merinos, with differences in meat quality between breeds.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Effects of the concrete crack on radiation shielding in spent fuel dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Min

    2007-02-15

    rays in a concrete structure with collinear cracks. In the case of abnormal state, the effect of spalling due to reinforcement corrosion on the shielding performance is estimated by numerical analysis. And because dry storage system for PWR is not decided yet, these effects applied to MACSTOR-400. In case that cover thickness is under 10cm, it is estimated that spalling will be formed. Nevertheless, if the thickness of wall in MACSTOR-400 is 100cm, it is estimated that surface dose rate will be maintained below the radiation safety standards. Dose rate at Wolsung site boundary where MACSTOR-400 is located regarding the effect of concrete cracking was estimated. This estimation is based on experimental results as mentioned above. This estimation is based on experimental results. In order to estimate dose rate at site boundary, the radiological effect of other facility in same site and the effect due to arrangement is considered. It is estimated that the dose rate increasing due to cracking is up to 25%. It seems that the additional shielding effect due to the 'mutual shielding effect' will be expected about 55-60%. It is estimated that the contribution from NPPs to dose rate are the most dominant. Concrete silo has surface dose rate lower than that of MACSTOR, however, because the number of concrete silo is much more than MACSTOR, it is estimated that contribution from concrete silo is more that that of MACSTOR-400. Although operation time of NPPs, it is estimated that dose rate is maintained under criteria, regardless of crack formation. The results of this study can be applied to assess the surface dose rate of concrete shield for collinear crack, and not irregular cracks with bends. However, they may be used to establish a standard for radiological safety in shielding structures, especially shielding margin. This study does not consider buildup factor. In order to be used in the industrial fields, a study on the effect of cracking on buildup is required

  4. 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.

  5. 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.

  6. 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.

  7. Dual purpose microalgae-bacteria-based systems that treat wastewater and produce biodiesel and chemical products within a biorefinery.

    Science.gov (United States)

    Olguín, Eugenia J

    2012-01-01

    Excess greenhouse gas emissions and the concomitant effect on global warming have become significant environmental, social and economic threats. In this context, the development of renewable, carbon-neutral and economically feasible biofuels is a driving force for innovation worldwide. A lot of effort has been put into developing biodiesel from microalgae. However, there are still a number of technological, market and policy barriers that are serious obstacles to the economic feasibility and competitiveness of such biofuels. Conversely, there are also a number of business opportunities if the production of such alternative biofuel becomes part of a larger integrated system following the Biorefinery strategy. In this case, other biofuels and chemical products of high added value are produced, contributing to an overall enhancement of the economic viability of the whole integrated system. Additionally, dual purpose microalgae-bacteria-based systems for treating wastewater and production of biofuels and chemical products significantly contribute to a substantial saving in the overall cost of microalgae biomass production. These types of systems could help to improve the competitiveness of biodiesel production from microalgae, according to some recent Life Cycle Analysis studies. Furthermore, they do not compete for fresh water resources for agricultural purposes and add value to treating the wastewater itself. This work reviews the most recent and relevant information about these types of dual purpose systems. Several aspects related to the treatment of municipal and animal wastewater with simultaneous recovery of microalgae with potential for biodiesel production are discussed. The use of pre-treated waste or anaerobic effluents from digested waste as nutrient additives for weak wastewater is reviewed. Isolation and screening of microalgae/cyanobacteria or their consortia from various wastewater streams, and studies related to population dynamics in mixed cultures

  8. Dual purpose system that treats anaerobic effluents from pig waste and produce Neochloris oleoabundans as lipid rich biomass.

    Science.gov (United States)

    Olguín, Eugenia J; Castillo, Omar S; Mendoza, Anilú; Tapia, Karla; González-Portela, Ricardo E; Hernández-Landa, Víctor J

    2015-05-25

    Dual purpose systems that treat wastewater and produce lipid rich microalgae biomass have been indicated as an option with great potential for production of biodiesel at a competitive cost. The aim of the present work was to develop a dual purpose system for the treatment of the anaerobic effluents from pig waste utilizing Neochloris oleoabundans and to evaluate its growth, lipid content and lipid profile of the harvested biomass and the removal of nutrients from the media. Cultures of N. oleoabundans were established in 4 L flat plate photobioreactors using diluted effluents from two different types of anaerobic filters, one packed with ceramic material (D1) and another one packed with volcanic gravel (D2). Maximum biomass concentration in D1 was 0.63 g L(-1) which was significantly higher than the one found in D2 (0.55 g L(-1)). Cultures were very efficient at nutrient removal: 98% for NNH4(+) and 98% for PO4(3-). Regarding total lipid content, diluted eflluents from D2 promoted a biomass containing 27.4% (dry weight) and D1 a biomass containing 22.4% (dry weight). Maximum lipid productivity was also higher in D2 compared to D1 (6.27±0.62 mg L(-1) d(-1) vs. 5.12±0.12 mg L(-1) d(-1)). Concerning the FAMEs profile in diluted effluents, the most abundant one was C18:1, followed by C18:2 and C16:0. The profile in D2 contained less C18:3 (linolenic acid) than the one in D1 (4.37% vs. 5.55%). In conclusion, this is the first report demonstrating that cultures of N. oleoabundans treating anaerobic effluents from pig waste are very efficient at nutrient removal and a biomass rich in lipids can be recovered. The maximum total lipid content and the most convenient FAMEs profile were obtained using effluents from a digester packed with volcanic gravel. Copyright © 2015 Elsevier B.V. All rights reserved.

  9. 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.

  10. 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.

  11. 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)

  12. Evaluation of Dual-purpose Cowpea (Vigna unguiculata (L. Walp. Varieties for Grain and Fodder Production at Shika, Nigeria

    Directory of Open Access Journals (Sweden)

    Omokanye, AT.

    2003-01-01

    Full Text Available A three-year field study of eight new and one check dual-purpose cowpea varieties was carried out to evaluate their grain and fodder production potential. Germination and seedling establishment were both high and greater than 80%.Mean dry fodder and seed yields varied from 1,262 to 3,598 kg/ha and 528 to 1,149 kg/ha respectively, with varieties IAR 4/48/15-1, IAR 72 and TVU 12349 retaining larger amounts (> 50% of fresh green leaves at pod harvest during the dry season. Crude protein (CP content of fodder averaged between 15.2 and 21.6%. There were more pods/plant for varieties IAR 4/48/15-1, IAR 7/180-4-5 and TVU 12349. 100-seed weight was highest with IT89KD-288 and Kananado (check. Fodder yield, pods/plant and leaf content were moderately correlated with seed yield. Results showed that varieties TVU 12349, IT89KD-288, IAR 2/180/4-12 and IAR 4/48/15-1 appeared suitable for both fodder and grain production. The use of appropriate cowpea varieties to enhance farmer income in an integrated production system is suggested.

  13. Economic analysis of alternative nutritional management of dual-purpose cow herds in central coastal Veracruz, Mexico.

    Science.gov (United States)

    Absalón-Medina, Victor Antonio; Nicholson, Charles F; Blake, Robert W; Fox, Danny Gene; Juárez-Lagunes, Francisco I; Canudas-Lara, Eduardo G; Rueda-Maldonado, Bertha L

    2012-08-01

    Market information was combined with predicted input-output relationships in an economic analysis of alternative nutritional management for dual-purpose member herds of the Genesis farmer organization of central coastal Veracruz, Mexico. Cow productivity outcomes for typical management and alternative feeding scenarios were obtained from structured sets of simulations in a companion study of productivity limitations and potentials using the Cornell Net Carbohydrate and Protein System model (Version 6.0). Partial budgeting methods and sensitivity analysis were used to identify economically viable alternatives based on expected change in milk income over feed cost (change in revenues from milk sales less change in feed costs). Herd owners in coastal Veracruz have large economic incentives, from $584 to $1,131 in predicted net margin, to increase milk sales by up to 74% across a three-lactation cow lifetime by improving diets based on good quality grass and legume forages. This increment is equal to, or exceeds, in value the total yield from at least one additional lactation per cow lifetime. Furthermore, marginal rates of return (change in milk income over feed costs divided by change in variable costs when alternative practices are used) of 3.3 ± 0.8 indicate clear economic incentives to remove fundamental productivity vulnerabilities due to chronic energy deficits and impeded growth of immature cows under typical management. Sensitivity analyses indicate that the economic outcomes are robust for a variety of market conditions.

  14. Effects of Dry Storage and Resubmersion of Oysters on Total Vibrio vulnificus and Total and Pathogenic (tdh+/trh+) Vibrio parahaemolyticus Levels.

    Science.gov (United States)

    Kinsey, Thomas P; Lydon, Keri A; Bowers, John C; Jones, Jessica L

    2015-08-01

    Vibrio vulnificus (Vv) and Vibrio parahaemolyticus (Vp) are the two leading causes of bacterial illnesses associated with raw shellfish consumption. Levels of these pathogens in oysters can increase during routine antifouling aquaculture practices involving dry storage in ambient air conditions. After storage, common practice is to resubmerge these stored oysters to reduce elevated Vv and Vp levels, but evidence proving the effectiveness of this practice is lacking. This study examined the changes in Vv and in total and pathogenic (thermostable direct hemolysin gene and the tdh-related hemolysin gene, tdh+ and trh+) Vp levels in oysters after 5 or 24 h of dry storage (28 to 32°C), followed by resubmersion (27 to 32°C) for 14 days. For each trial, replicate oyster samples were collected at initial harvest, after dry storage, after 7 days, and after 14 days of resubmersion. Oysters not subjected to dry storage were collected and analyzed to determine natural undisturbed vibrio levels (background control). Vibrio levels were measured using a most-probable-number enrichment followed by real-time PCR. After storage, vibrio levels (excluding tdh+ and trh+ Vp during 5-h storage) increased significantly (P 0.1) from levels in background oysters after 14 days of resubmersion, regardless of dry storage time. These data demonstrate that oyster resubmersion after dry storage at elevated ambient temperatures allows vibrio levels to return to those of background control samples. These results can be used to help minimize the risk of Vv and Vp illnesses and to inform the oyster industry on the effectiveness of routine storing and resubmerging of aquaculture oysters.

  15. 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...

  16. The application of mature dry storage technology and remote handling robotics to nuclear plant extension, clean-up and decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Blackwell, W.C. [GEC ALSTHOM Engineering Systems Ltd., Leicester (United Kingdom)

    1997-08-01

    This paper reviews a mature dry storage technology developed by GEC ALSTHOM Engineering Systems Limited (GAES) which offers a passive, economical and licensable method of providing irradiated fuel storage capacity at operational nuclear power stations. The evolution of the modular vault dry store (MVDS) technology has taken place over 25 years of operational experience, culminating in a product which meets all of the concerns of licensing authorities regarding safety and fuel integrity. The application of remote handling robotics to nuclear fuel and active component handling as a routine process rather than as an intervention technique is also reviewed. The growth of the application of this technology is governed by several factors which include: statutory requirements, safety assurance, risk reduction and economic pressures. The availability of a mature MVDS technology with an evolving process-capable robotics technology opens up opportunities for exploring proven UK products. (Author).

  17. Feasibility study in aspect of thermal integrity on the dry storage expansion options for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.; Song, M. J. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, K. S. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    In order to expand the capability of the CANDU spent fuel dry storage facilities of the at Wolsong, the alternative concepts based on MACSTOR are suggested to replace with existing concrete silo of Wolsong. For this, the feasibility of its design changes from original MACSTOR is examined in term of heat transfer and thermal hydraulic. In this study, the configuration of the module was conceptually changed from its original 2 rows to 3 and 4 rows for review. Under normal operation, the results of heat transfer and thermal hydraulic shows that storage module can feasibly accomodate four rows of storage cylinders within allowable range in terms of maximum allowable temperature of the fuel basket.

  18. Dual-purpose laser irradiation and perfusion testing system for in-vitro experiments using cultured trabecular meshwork endothelial cells

    Science.gov (United States)

    Rivera, Brian K.; Roberts, Cynthia J.; Weber, Paul A.

    1998-06-01

    The means by which Argon laser trabeculoplasty (ALT) lowers intraocular pressure (IOP) is a matter of debate. Mechanical and biological laser-tissue interaction theories have been proposed. To investigate the effect laser irradiation has upon the aqueous outflow facility of trabecular meshwork (TM) cells, a suitable in-vitro model is required. Therefore the purpose of this study was to design, construct, and validate a laser irradiation and perfusion testing apparatus. The system was designed to utilize cultured TM cells seeded onto filter supports. Outflow facility will be quantified by calculating the hydraulic conductivity of the monolayer. An appropriate filter support was located, and its perfusion characteristics determined using water. Afterwards, the steady state perfusion flow rate of the filter was ascertained to be 0.096 plus or minus 0.008 ml/min when culture medium is used. Following these tests a single, baseline perfusion experiment was conducted using a TM cell monolayer. Analysis of the data produced a baseline hydraulic conductivity of 0.673 plus or minus 0.076 (mu) l/min/mm Hg/cm2, well within the range found in previous reports. A dual purpose, in vitro-cellular perfusion and laser irradiation testing apparatus has been developed, tested and validates using known baseline cellular perfusion and laser irradiation testing apparatus has been developed, tested, and validated using known baseline cellular perfusion values. Future experiments will be conducted to verify these initial findings, and further experiments will be conducted using Argon laser irradiation. The response of the TM cell monolayer will then be compared to the baseline figures.

  19. Effects of Fleckvieh on the Performance of Dual-Purpose Cattle Improved by Hybridization with Local Beef Cows

    Institute of Scientific and Technical Information of China (English)

    Wang Linfeng; Yang Gaiqing; Lian Hongxia; Yan Ping; Liu Xian; Liu Yan; Wang Wenhui; Zhu Ruiguang; Li Ming; Yang Guoyu

    2015-01-01

    [Objective] The paper aimed to study the improved effect of local beef cattle with Fleckvieh cattle and provide theoretical basic data for local cattle industry in central plain agricultural area in China. [Method] With 500 local beef cattle as female parents and Fleckvieh cattle as male parents,hybridization improvement was conducted via artificial insemination. The growth performance,slaughter performance,milk performance and milk components of F1 and F2 hybrids were measured. [Result] The birth body weights of F2 were significantly higher than those of local beef cow,but there was no remarkable difference between F1 and local beef cow or F1 and F2. The growth rates of F1 and F2 at different stages were higher than those of local beef cattle. The slaughter performance,such as carcass weight( P < 0. 05),dressing percentage,net meat rate( P < 0. 05),marbling score of F1 and F2 were higher than those of local beef cow. Milk production performance,such as actual milk yield,305 d corrected milk yield and 4% standard milk yield of F2 were signally higher than those of F1 and local beef cattle( P< 0. 05),and F1 was markedly higher than local beef cattle( P < 0. 05). For milk composition,although milk fat percentage,milk protein rate,lactose rate and total solids( TS) of F1 and F2 were slightly lowered compared with local beef cattle at varying degrees,they were still at high levels compared with Holstein cows.[Conclusion]Fleckvieh cattle,as a male parent,can significantly improved growth performance,slaughter performance and milk performance of offsprings. It would also increases the economic efficiency of local beef cattle by higher quality and price,as well as changing production model from beef to dual purpose of beef and milk.

  20. 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......

  1. 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''...

  2. 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

    could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ ISFSIs are adequately managed for the period of extended operation. To date, all of the ISFSIs located across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years; three ISFSIs (Surry, H.B. Robinson and Oconee) have received their renewed licenses for 20 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. This report examines issues related to managing aging effects on the SSCs in DCSSs/ISFSIs for extended long-term storage and transportation of used fuels, following an approach similar to that of the Generic Aging Lessons Learned (GALL) report, NUREG-1801, for the aging management and license renewal of nuclear power plants. The report contains five chapters and an appendix on quality assurance for aging management programs for used-fuel dry storage systems. Chapter I of the report provides an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927. Chapter II contains definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms. Chapter III and Chapter IV contain generic TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs described in Chapter V. The summary descriptions and tabulations of evaluations of AMPs and TLAAs for the SSCs that are important to safety in Chapter V include DCSS designs (i.e., NUHOMS{reg_sign}, HI-STORM 100, Transnuclear (TN) metal cask, NAC International S/T storage cask, ventilated storage cask (VSC-24), and the Westinghouse MC-10 metal dry storage cask) that have been and continue to be used by utilities across the country for

  3. 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

    could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ ISFSIs are adequately managed for the period of extended operation. To date, all of the ISFSIs located across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years; three ISFSIs (Surry, H.B. Robinson and Oconee) have received their renewed licenses for 20 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. This report examines issues related to managing aging effects on the SSCs in DCSSs/ISFSIs for extended long-term storage and transportation of used fuels, following an approach similar to that of the Generic Aging Lessons Learned (GALL) report, NUREG-1801, for the aging management and license renewal of nuclear power plants. The report contains five chapters and an appendix on quality assurance for aging management programs for used-fuel dry storage systems. Chapter I of the report provides an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927. Chapter II contains definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms. Chapter III and Chapter IV contain generic TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs described in Chapter V. The summary descriptions and tabulations of evaluations of AMPs and TLAAs for the SSCs that are important to safety in Chapter V include DCSS designs (i.e., NUHOMS{reg_sign}, HI-STORM 100, Transnuclear (TN) metal cask, NAC International S/T storage cask, ventilated storage cask (VSC-24), and the Westinghouse MC-10 metal dry storage cask) that have been and continue to be used by utilities across the country for

  4. 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.

  5. 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.

  6. Seismic Structure-Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Kim, Sung Hwan; Yang, Ke Hyung; Lee, Heung Young; Cho, Chun Hyung [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su [KONES Corporation, Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside the concrete module consists of 40 storage cylinders accommodating ten 60-bundle dry storage baskets, which are suspended from the top slab and eventually restrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module shall be by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants, except for local site characteristics required for soilstructure interaction (SSI) analysis. It is required for the structural integrity to fulfill the licensing requirements. As per USNRC SRP Section 3.7.2, it shall be reviewed how to consider the phenomenon of coupling of the dynamic response of adjacent structures through the soil, which is referred to as structure-soil-structure interaction (SSSI). The presence of closely spaced multiple structural foundations creates coupling between the foundations of individual structures . Some observations of the actual seismic response of structures have indicated that SSSI effects do exist, but they are generally secondary for the overall structural response motions. SSSI effects, however, may be important for a relatively small structure which is to be close to a relatively large structure, while they may be generally neglected for overall structural response of a large massive structure, such as nuclear power plant. As such the scope of the present paper is to carry out a seismic SSSI analysis in case of the MACSTOR/KN- 400 module, in order to investigate whether or not SSSI effect shall be included in the overall seismic

  7. Dual-Purpose Bioreactors to Monitor Noninvasive Physical and Biochemical Markers of Kidney and Liver Scaffold Recellularization.

    Science.gov (United States)

    Uzarski, Joseph S; Bijonowski, Brent M; Wang, Bo; Ward, Heather H; Wandinger-Ness, Angela; Miller, William M; Wertheim, Jason A

    2015-10-01

    Analysis of perfusion-based bioreactors for organ engineering and a detailed evaluation of physical and biochemical parameters that measure dynamic changes within maturing cell-laden scaffolds are critical components of ex vivo tissue development that remain understudied topics in the tissue and organ engineering literature. Intricately designed bioreactors that house developing tissue are critical to properly recapitulate the in vivo environment, deliver nutrients within perfused media, and monitor physiological parameters of tissue development. Herein, we provide an in-depth description and analysis of two dual-purpose perfusion bioreactors that improve upon current bioreactor designs and enable comparative analyses of ex vivo scaffold recellularization strategies and cell growth performance during long-term maintenance culture of engineered kidney or liver tissues. Both bioreactors are effective at maximizing cell seeding of small-animal organ scaffolds and maintaining cell survival in extended culture. We further demonstrate noninvasive monitoring capabilities for tracking dynamic changes within scaffolds as the native cellular component is removed during decellularization and model human cells are introduced into the scaffold during recellularization and proliferate in maintenance culture. We found that hydrodynamic pressure drop (ΔP) across the retained scaffold vasculature is a noninvasive measurement of scaffold integrity. We further show that ΔP, and thus resistance to fluid flow through the scaffold, decreases with cell loss during decellularization and correspondingly increases to near normal values for whole organs following recellularization of the kidney or liver scaffolds. Perfused media may be further sampled in real time to measure soluble biomarkers (e.g., resazurin, albumin, or kidney injury molecule-1) that indicate degree of cellular metabolic activity, synthetic function, or engraftment into the scaffold. Cell growth within bioreactors is

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. Chinese and International Situation, Progresses and Perspectives of Breeding Strategies in Dual Purpose Cattle%国内外兼用牛现状及育种理论研究进展

    Institute of Scientific and Technical Information of China (English)

    马秋萌; 秦春华; 王雅春; 史远刚; 张胜利; 张沅

    2013-01-01

    Dual purpose cattle is famous by its elite performance in both milk and meat production.This study first reviewed the genetic resources of dual purpose cattle in China and major dual purpose cattle breeds internationally,then summarized the development of breeding system and breeding strategies applied in dual purpose cattle breeding,to provide reference for drawing up breeding scheme for dual purpose cattle in China.%兼用牛亦称乳肉兼用牛,是兼具较高产奶性能和优秀产肉性能的牛品种.本文就国内外兼用牛的发展现状、育种体系和育种规划方法的研究进展做简要综述,为制定我国兼用牛育种规划提供参考.

  13. Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

    2004-07-01

    The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349.

  14. 双重功能图像水印算法%An image watermarking algorithm with dual purpose

    Institute of Scientific and Technical Information of China (English)

    叶天语

    2011-01-01

    An image watermarking algorithm with dual-purpose is proposed,which can achieve both content authentication and copyright protection. Firstly, the original image was divided into non-overlapping blocks,and each block was conducted with singular value decomposition (SVD). A robust zero-watermarking .sequence was produced by judging the numerical relationship between each block's biggest singular value and the average of biggest singular values from all blocks. Afterwards,the original image was divided into non-overlapping blocks,and each block was transformed with discrete cosine transformation (DCT). Several high frequency coefficients were adjusted to establish linear numerical relationship between two different DCT coefficients from the same block,and the modified image was received after inverse discrete cosine transformation (IDCT). Content authentication was accomplished by judging whether there was a linear numerical relationship between two relevant DCT coefficients from each block of the tampered image based on DCT's reversibility. Copyright was identified by the similarity between the original robust zero-watermarking sequence and the robust zero-watermarking sequence extracted from the attacked image. The experimental results show that the algorithm has good invisibility,and can achieve both content authentication and copyright protection.%提出一种能同时实现内容认证和版权保护双重功能的图像水印算法.首先,对原始图像进行分块奇异值分解(SVD),计算所有子块最大奇异值的均值,通过比较各子块的最大奇异值与所有子块最大奇异值的均值间的数值关系产生鲁棒零水印序列.然后,对原始图像进行分块离散余弦变换(DCT).调整图像子块DCT高频系数的数值大小,建立同一子块两个不同DCT系数间数值的线性关系,通过逆DCT(IDCT)得到系数调整后的图像.内容认证时,利用DCT的可逆性,通过判断遭篡改图像每个子块两个相应DCT

  15. A design report for DFDF radioactive waste storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Kim, J. H.; Park, J. J.; Yang, M. S.; Lee, J. C.; Seo, K. S

    2004-01-01

    The fabrication of DUPIC fuel powder, pellet and fuel rod has been conducted since the early 2000 at DFDF hot cell, together with their fabrication characterization experiment. The equipment that were used in the experiment, if necessary, have to be disassembled to be transported to the storage facility according to the appropriate management procedures. Therefore, specially designed dedicated storage cask is needed for the waste of large volume such as contaminated equipment and disassembled parts from the equipment. The storage cask, 'A' type transportation cask, has a cylindrical shape and consists of body and lid. It also has a dual cell structure. The cells are made of stainless steel and the lead is molded into the space between cells as a gamma radiation protector. The storage cask is 1,145mm high and weighs 3.5 ton. Its inner and outer diameter are 1,080mm and 1,210mm, respectively. It is designed to maintain at 0.3kg/cm{sup 2} to prevent possible leakage during the storage.

  16. 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.

  17. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  18. 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.

  19. 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.

  20. Quantitation of selected terpenoids and mercaptans in the dual-purpose hop varieties Amarillo, Citra, Hallertau Blanc, Mosaic, and Sorachi Ace.

    Science.gov (United States)

    Cibaka, Marie-Lucie Kankolongo; Gros, Jacques; Nizet, Sabrina; Collin, Sonia

    2015-03-25

    Free terpenoids and both free and bound polyfunctional thiols were investigated in five selected dual-purpose hop cultivars. Surprisingly, the dual-purpose Sorachi Ace variety was found to contain higher amounts of farnesene (2101 mg/kg) than aromatic hops such as Saaz but only traces of 3-methylbutylisobutyrate, a compound that usually distinguishes all bitter varieties. All five cultivars investigated here showed an exceptional citrus-like potential explained by either monoterpenic alcohols or polyfunctional thiols. Among the monoterpenic alcohols, β-citronellol at concentrations above 7 mg/kg distinguished Amarillo, Citra, Hallertau Blanc, Mosaic, and Sorachi Ace from Nelson Sauvin and Tomahawk, two previously investigated dual-purpose hops, while linalool (312 mg/kg) and geraniol (211 mg/kg) remained good discriminating compounds for Nelson Sauvin and Tomahawk, respectively. Regarding polyfunctional thiols, higher amounts of 3-sulfanylhexyl acetate (27 μg/kg) characterized the Citra variety. Free 4-sulfanyl-4-methylpentan-2-one proved discriminant for Sorachi Ace, while the bound form is predominant in Nelson Sauvin. On the other hand, an S-conjugate of 3-sulfanylhexan-1-ol was found in Sorachi Ace at levels not far from those previously reported for Cascade, although the free form was undetected here. Both free and bound grapefruit-like 3-sulfanyl-4-methylpentan-1-ol (never evidenced before the present work) emerged as discriminating compounds for the Hallertau Blanc variety. The apotryptophanase assay also allowed us to evidence for the first time an S-conjugate of 2-sulfanylethan-1-ol.

  1. 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.

  2. 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

  3. 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).

  4. 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...

  5. 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.

  6. 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.

  7. 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.

  8. Prevalence of and risk factors for bovine respiratory syncytial virus (BRSV) infection in non-vaccinated dairy and dual-purpose cattle herds in Ecuador.

    Science.gov (United States)

    Saa, Luis Rodrigo; Perea, Anselmo; Jara, Diego Vinicio; Arenas, Antonio José; Garcia-Bocanegra, Ignacio; Borge, Carmen; Carbonero, Alfonso

    2012-10-01

    A cross-sectional study was carried out to determine the seroprevalence and risk factors associated with bovine respiratory syncytial virus (BRSV) infection in non-vaccinated dairy and dual-purpose cattle herds from Ecuador. A total of 2,367 serum samples from 346 herds were collected from June 2008 to February 2009. A questionnaire, which included variables related to cattle, health, management measures, and the environment, was filled out in each herd. Presence of antibodies against BRSV was analyzed using a commercial indirect ELISA test. A logistic regression model was used to determine risk factors associated with BRSV at herd level. The individual seroprevalence against BRSV in non-vaccinated herds in Ecuador was 80.48% [1,905/2,367; 95% confidence interval (CI) = 78.9-82.1]. The herd prevalence was 91.3% (316/346; 95% CI = 88.3-94.3), and the intra-herd prevalence ranged between 25% and 100% (mean, 90.47%). The logistic regression model showed that the existence of bordering cattle farms, the dual-purpose farms, and the altitude of the farm (more than 2,338 m above sea level) were risk factors associated with BRSV infection. This is the first study about BRSV prevalence in Ecuador. It shows the wide spread of the BRSV infection in the country. The risk factors found will help to design effective control strategies.

  9. Bonner sphere neutron spectrometry at spent fuel casks

    CERN Document Server

    Rimpler, A

    2002-01-01

    For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon-neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locat...

  10. 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.

  11. 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.

  12. 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

  13. Development of a probabilistic safety assessment framework for an interim dry storage facility subjected to an aircraft crash using best-estimate structural analysis

    Energy Technology Data Exchange (ETDEWEB)

    Almomani, Belal; Jang, Dong Chan [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Kang, Hyun Gook [Dept. of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy (United States)

    2017-03-15

    Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

  14. Estimates of power deposited via cesium/barium beta and gamma radiation captured in components of a Hanford cesium chloride capsule and by components of overpacked capsules placed in an interim dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Roetman, V.E., Westinghouse Hanford

    1996-12-23

    The deposition of power in Hanford cesium chloride capsules and in the components of design concepts for overpacking and interim storage were determined as requested (Randklev, 1996a). The power deposition results from the selective capture of gamma and beta radiation coming from the decay of the 137CS isotope in the CsCl contained in the capsules. The following three cases were analyzed: (a) a single CsCl capsule, (b) an overpack containing eight CsCl capsules, and (c) an infinite square array of such overpacks as placed in tubes of a interim dry storage facility. The power deposition was expressed as watts per gram for each of the respective physical design components in these three cases. Per the analyses request and guidance (Randklev 1996a), the primary analysis objective was to characterize, for each case, the power deposition across the radial cross-section at the expected axial position of maximum deposition. As requested, this primary part of the analysis work was done using choices for component dimension and material properties that would reasonably characterize the maximum deposition profile across the salt (CsCl) and the inner capsule barrier of the double walled metal capsule system used to construct the Hanford capsules. The secondary objective was to further evaluate the deposition behavior relative to the influence of axial position. The guidance (Randklev 1996a) also requested 1797 an analysis case that involved a lag-storage pit in a hot-cell, in which a cylindrical metal basket from a transportation cask would be used to position several capsules in the lag-storage pit. Although the basic model for the lag storage concept evaluation was essentially completed by the end of FY-96, the analysis was not run because of the need to prioritize and limit the work scope due to funding limitations for FY-97. The specific purpose for performing the subject set of analyses (Randklev 1996a) is to obtain power deposition values (i.e., per the decay of T37cs

  15. 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

  16. 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...

  17. 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

  18. 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.

  19. A Dual-Purpose Ion-Accelerator for Nuclear-Reaction-Based Explosives-and SNM-Detection in Massive Cargo

    CERN Document Server

    Goldberg, M B; Vartsky, D; Bar, D; Böttger, R; Brandis, M; Bromberger, B; Feldman, G; Friedman, E; Heflinger, D; Lauck, R; Löb, S; Maier-Komor, P; Mardor, I; Mor, I; Speidel, K -H; Tittelmeier, K; Weierganz, M

    2010-01-01

    A dual-purpose ion-accelerator concept, capable of serving as radiation source in a versatile, nuclear-reaction-based cargo inspection system, is presented. The system will automatically and reliably detect small, operationally-relevant quantities of concealed explosives and special nuclear materials (SNM). It will be cost-effective, employing largely-common hardware, but different reactions/DAQ-modes. Typical expected throughput is 10-20 aviation containers/hr. PACS: 25.20.Dc; 25.40.Ny; 27.20.+n; 29.27.Fh; 79.77.+g; 89.20.Bb; 89.20.Dd Keywords: Cargo inspection; Nuclear-reaction-based methods; Explosives detection; SNM detection

  20. 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.

  1. Effects of extended dry storage of powdered infant milk formula on susceptibility of Enterobacter sakazakii to hot water and ionizing radiation.

    Science.gov (United States)

    Osaili, Tareq M; Al-Nabulsi, Anas A; Shaker, Reyad R; Ayyash, Mutamed M; Olaimat, Amin N; Al-Hasan, Ashraf S Abu; Kadora, Khaled M; Holley, Richard A

    2008-05-01

    Infant milk formula has been identified as a potential source of Enterobacter sakazakii, which has been implicated in neonatal meningitis and necrotizing enterocolitis. This study was undertaken to determine whether the length of E. sakazakii storage in powdered infant milk formula (PIMF) affected the ability of the pathogen to survive subsequent reconstitution of the powder with hot water or treatment with gamma radiation. Five E. sakazakii strains were mixed individually with PIMF and kept for up to 12 months at 25 degrees C. After storage PIMF was reconstituted with water at 60 to 100 degrees C or was exposed to radiation. Without any treatment secondary to drying, E. sakazakii counts decreased Dry storage decreased thermal resistance but increased resistance of E. sakazakii to ionizing radiation in PIMF. Reconstitution of contaminated powder with water at 70 degrees C after 1 month of dry storage reduced E. sakazakii viability slightly, > 2 log/g, and after powder was stored for 12 months all E. sakazakii strains were eliminated. In contrast, desiccation substantially increased the resistance of E. sakazakii strains to ionizing radiation. Although the D-value for E. sakazakii IMF1 following overnight storage in PIMF was 0.98 kGy, > 4 kGy was required to kill 1.5 log/g of the same strain that had survived 12 months in dry PIMF. Results suggested that low-dose irradiation will more effectively eliminate E. sakazakii from PIMF if the treatment is applied shortly after PIMF manufacture.

  2. Exploration and application of wet-discharging dry storage of phosphogypsum%磷石膏湿排干堆的探索和应用

    Institute of Scientific and Technical Information of China (English)

    王光明; 曹绪勇

    2011-01-01

    Based on the disadvantages of wet-discharging wet storage and dry-discharging dry storage phosphogypsum commonly used in compound fertilizer enterprise, the wet-discharging dry storage of phosphogypsum is put forward, and its process operation and advantages are introduced, which evades effectively the high threshold and risk of tailings reservoir, improves the environment, decreases the cost of operation and maintenance, has the obvious superiority.%针对目前磷复肥企业磷石膏堆存常用的湿排湿堆及千排干堆方法所存在的诸多弊端,提出采用湿排干堆,并介绍其工艺运行情况及优势,有效规避了建设尾矿库的高门槛、高风险,改善了现场环境,降低了运行维护费用,优势明显。

  3. Reuse inspection refort of the spent fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. W.; Seo, K. S.; Ku, J. H.; Lee, J. C.; Bang, K. S.; Min, D. K. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-04-01

    This is the contract result report performed by KAERI under the contract with KPS for the reuse inspection of the KSC-4 No. 2 cask to receive the license for the reuse of next 5 years. According to the revision of the atomic regulations, all type B package should receive and pass the reuse inspection for every 5 years. This report contains the summary of the reuse inspection project, the details of the inspection methods and evaluation criteria, the documents which submitted to the KINS and the license approved by the KINS. 1 tabs. (Author)

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. Material specification and quality control program for ductile iron spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Rehmer, B.; Frenz, H.; Weidlich, S.; Kuehn, H.D.

    1995-12-31

    In the process of testing spent fuel casks, BAM is gaining a lot of relevant data regarding the quality level of Ductile Cast Iron (DCI). This paper discusses the basic parameters governing the material behavior of ferritic and ferritic-pearlitic DCI and reviews the development of cask quality over the last years. The effect of microstructure and sample size on the fracture toughness of DCI is discussed. The results of a test program show the prominent effect of pearlite content and graphite nodule structure in the mechanical and fracture toughness characteristics of DCI. This observation is important for quality assurance programs for shipping and storage casks of radioactive materials.

  11. 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)

  12. 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)

  13. Design of Bunk Bed with Dual-purpose Ladder%学生双层床两用梯的设计

    Institute of Scientific and Technical Information of China (English)

    张西珠; 耿振香

    2016-01-01

    Based on an analysis of the bunk bed in a current student’s dormitory, a dual-purpose ladder of the bunk bed for students is designed to diversity the function of the ladder. This dual-propose ladder is more than a common one, it can be used as a table after being folded up by sliding and flipping. The ladder’s function can be improved as well as a guarantee of security. The designed ladder not only change the understanding of a bunk bed ladder, but also highlight its perfect innovation features.%通过对目前宿舍学生双层床梯子功能的分析,设计出一种新型的学生双层床两用梯。本文不仅介绍了梯子通过滑动和翻转,使两用梯在完成正常梯子功能后,又可折叠当作书桌使用功能;在保证安全性的同时,完成了结构性设计。本设计具有创新特性,让双层床的梯子功能实现了多样化。

  14. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  15. 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.

  16. 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.

  17. Generating thermal energy and electric power by yourself. Heating system modernization with a mini dual-purpose power plant; Waerme und Strom selbst erzeugen. Heizungsmodernisierung mit Mini-BHKW

    Energy Technology Data Exchange (ETDEWEB)

    Meissner, H. [Powerplus Technologies GmbH, Gera (Germany)

    2008-03-15

    Mini dual-purpose power plants offer many advantages to the customer by means of reduced primary energy consumoption, reduced CO{sub 2} emission and considerable savings of energy costs. The following contribution describes the balance in the one-family house of the Gross family and what heating systems master craftsman Keijo Sprogoetook into consideration when he executed this modernization task. (orig.)

  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. 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.

  1. 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.

  2. COBRA-SFS modifications and cask model optimization

    Energy Technology Data Exchange (ETDEWEB)

    Rector, D.R.; Michener, T.E.

    1989-01-01

    Spent-fuel storage systems are complex systems and developing a computational model for one can be a difficult task. The COBRA-SFS computer code provides many capabilities for modeling the details of these systems, but these capabilities can also allow users to specify a more complex model than necessary. This report provides important guidance to users that dramatically reduces the size of the model while maintaining the accuracy of the calculation. A series of model optimization studies was performed, based on the TN-24P spent-fuel storage cask, to determine the optimal model geometry. Expanded modeling capabilities of the code are also described. These include adding fluid shear stress terms and a detailed plenum model. The mathematical models for each code modification are described, along with the associated verification results. 22 refs., 107 figs., 7 tabs.

  3. 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.

  4. Specific outcomes of the research on the spent fuel long-term evolution in interim dry storage and deep geological disposal

    Science.gov (United States)

    Ferry, C.; Poinssot, C.; Cappelaere, C.; Desgranges, L.; Jegou, C.; Miserque, F.; Piron, J. P.; Roudil, D.; Gras, J. M.

    2006-06-01

    This paper presents an overview of the main results of the French research on the long-term evolution of spent fuel. The behavior of the spent fuel rods in the various conditions likely to be encountered during dry storage and deep geological disposal, i.e., in a closed system, in air and in water were investigated. It appears that in a closed system the effects of helium production on the mechanical stability of grain boundaries remain the major unanswered question. In air, microscopic characterization of the UO2 oxidation leads to introduce a new phase in the classical oxidation scheme. The limiting step assumption on which the oxidation kinetics are based is only partially valid. In water, the effect of the alpha radiolysis which accelerates UO2 dissolution was demonstrated for anoxic conditions. However this effect could be counteracted by the environmental conditions, such as the presence of H2 produced by the container corrosion. The effects of the environmental parameters on the fuel matrix dissolution still need to be assessed.

  5. 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...

  6. 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.

  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. 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.

  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. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Mechanical and fracture behavior of nuclear fuel cladding in terms of transport and temporary dry storage; Comportamiento mecanio y en fractura de vainas de combustible nuclear en condiciones de transporte y almacenamiento temporal en seco

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Hervias, J.; Martin Rengel, M. A.; Gomez, F. J.

    2012-11-01

    In this work, the most relevant results of a research project on the mechanical and fracture behavior of cladding in transport and dry storage conditions are summarized. the project is being carried out at Universidad Politecnica de Madrid in collaboration with ENUSA, ENRESA and CSN. Non-irradiated cladding is investigated. The main objective is to determine a failure criterion of cladding as a function of hydrogen content, temperature and strain rate. (Author)

  16. Remote search for gamma sources at the dry storage unit 3A in Andreeva bay and the determination of their dose rates during activities for improvement to the radiological environment

    OpenAIRE

    Ivanov Kirill E.; Varnavin Anatoly P.; Korolev Andrey V.; Stepennov Boris S.; Sukhoruchkin Andrey K.; Teterin Yury A.; Teterin Anton Yu.; Kharitonov Vladimir V.; Selishev Valery A.; Fedoseenkov Alexander N.; Krasnoshekov Alexander N.; Kosnikov Alexander S.; Kostikov Dmitry A.

    2012-01-01

    The remotely controlled replacement of the concrete covering of the spent fuel dry storage unit 3A with a new iron horizontal biological shielding was carried out during works aimed at the improvement of the radiological environment at the NWC “SevRAO” - Branch of FSUE “RosRAO”, Andreeva Bay, Murmansk Region. Video control systems, a BROKK robotic manipulator, HIAB manipulator crane, gamma detectors of the ASCRO radiation monitoring system, and a CARTOGAM gamma camera were employed. A C...

  17. Remote search for gamma sources at the dry storage unit 3A in Andreeva bay and the determination of their dose rates during activities for improvement to the radiological environment

    Directory of Open Access Journals (Sweden)

    Ivanov Kirill E.

    2012-01-01

    Full Text Available The remotely controlled replacement of the concrete covering of the spent fuel dry storage unit 3A with a new iron horizontal biological shielding was carried out during works aimed at the improvement of the radiological environment at the NWC “SevRAO” - Branch of FSUE “RosRAO”, Andreeva Bay, Murmansk Region. Video control systems, a BROKK robotic manipulator, HIAB manipulator crane, gamma detectors of the ASCRO radiation monitoring system, and a CARTOGAM gamma camera were employed. A CARTOGAM gamma camera was used in all stages of the work involving high radiation levels for the remote location of the most dangerous gamma radiation sources and the evaluation of their dose rates. Gamma detectors of the ASCRO radiation monitoring system were located at several spots of the dry storage unit 3A in order to control the radiation situation. The use of the ASCRO and CARTOGAM has allowed us to avoid unauthorized exposure of the staff involved in the operations at the dry storage unit 3A site.

  18. 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.

  19. 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.

  20. Seismic and cask drop excitation evaluation of the tower shielding reactor

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N. (EQE, Inc., San Francisco, CA (USA); Oak Ridge National Lab., TN (USA); EQE, Inc., San Francisco, CA (USA))

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs.

  1. 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.

  2. Índices de seleção para um rebanho Caracu de duplo propósito Selection indices for a dual purpose breed Caracu

    Directory of Open Access Journals (Sweden)

    Sandra Aidar de Queiroz

    2005-06-01

    objective (H considered a dual purpose breed producing milk and weaning calves. The biological traits affecting income and expenses considered in H were: total milk yield, age at first calving, number of days open, weaning weight, and productive lifetime. Two economic selection indices were calculated. The selection criteria included in Index 1 were first lactation milk yield, number of days open from first to second calving, scrotal circumference adjusted to age and weight at 15 months and weaning weight (WW. Index 2 included the same traits but WW was switched by average daily gain from birth to weaning. The statistical analyses to estimate the (covariances and genetic and phenotypic parameters were performed by derivative-free restricted maximum likelihood method, using an animal model (uni and bi-trait. Records on animal performances, pedigrees and prices were provided by the breeder. The production costs and revenues of this herd were calculated from 1994 to 2000. The profit equation was set to an annual base using annual averages for number of animals per category, biological traits and prices. The economic value of each trait was obtained as the partial derivative of the profit function with respect to that trait (evaluated at the average for all other traits. Both indexes would improve selection response to H, but the economic selection index 1, including WW, would be a little more efficient for total selection response to H.

  3. 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

  4. 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...

  5. 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.

  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. 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.

  8. 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.

  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. 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

  11. Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

    Directory of Open Access Journals (Sweden)

    Rezaeian Mahdi

    2015-01-01

    Full Text Available Containment of a transport cask during both normal and accident conditions is important to the health and safety of the public and of the operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated. The contributions to the total activity from the four sources of gas, volatile, fines, and corrosion products are treated separately. These calculations are necessary to identify an appropriate leak test that must be performed on the cask and the results can be utilized as the source term for dose evaluation in the safety assessment of the cask.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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)

  18. Probabilistic Risk Assessment of Cask Drop Accident during On-site Spent Nuclear Fuel Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Jae Hyun; Christian, Robby; Momani, Belal Al; Kang, Hyun Gook [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are two ways to transfer the SNF from a site to other site, one is land transportation and the other is maritime transportation. Maritime transportation might be used because this way uses more safe route which is far from populated area. The whole transportation process can be divided in two parts: transferring the SNF between SNP and wharf in-Nuclear Power Plant (NPP) site by truck, and transferring the SNF from the wharf to the other wharf by ship. In this research, on-site SNF transportation between SNP and wharf was considered. Two kinds of single accident can occur during this type of SNF transportation, impact and fire, caused by internal events and external events. In this research, PRA of cask drop accident during onsite SNF transportation was done, risk to a person (mSv/person) from a case with specific conditions was calculated. In every 11 FEM simulation drop cases, FDR is 1 even the fuel assemblies are located inside of the cask. It is a quite larger value for all cases than the results with similar drop condition from the reports which covers the PRA on cask storage system. Because different from previous reports, subsequent impact was considered. Like in figure 8, accelerations which are used to calculate the FDR has extremely higher values in subsequent impact than the first impact for all SNF assemblies.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 3D analysis of thermo-fluid dynamics of a dry storage fuel container in stationary conditions; Analisis 3D de la termo-fluidodinamica de un contenedor de almacenamiento en seco de combustible en condiciones estacionarias

    Energy Technology Data Exchange (ETDEWEB)

    Penalva, J.; Feria, F.; Herranz, L. E.

    2012-07-01

    Dry storage containers must ensure the cooling of the fuel housing. Compliance with this requirement is of huge importance to preserve the integrity of spent fuel. In this sense, the thermo-fluid dynamics of containers is a point to consider in safety studies of this storage system. The aim of this work is to achieve a three-dimensional model of thermo-fluid dynamics of the HI-STORM 100S container using Fluent code. In addition to the fundamental characterization of the device, we have studied the impact of design variations associated with the input and output channels air. In the future, the model presented here will provide a basis for analysis of transient and accidental conditions.

  7. 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.

  8. Effects of Ensilaged Apple Pomace on Milk Yield and Quality of Simmental Hybrid Dual-Purpose Cattles%青贮苹果渣对西门塔尔牛产奶量及乳品质的影响

    Institute of Scientific and Technical Information of China (English)

    严平

    2016-01-01

    选取年龄、体重相近和首次产犊且处于泌乳后期(200d)的西门塔尔牛60头,采用完全随机区组设计分为6组,对照组饲喂由青贮玉米、苜蓿草粉、小麦秸和精料组成的基础日粮,试验组日粮中分别以5kg、10kg、15kg、20kg、25kg的青贮苹果渣替代等量青贮玉米。结果表明,以青贮苹果渣替代青贮玉米对西门塔尔牛的产奶量和乳品质均有一定程度的提高,替代50%时可显著提高牛的产奶量、乳脂率和乳蛋白含量(P0.05)。试验表明,青贮苹果渣替代一定比例的青贮玉米对西门塔尔牛的产奶量、牛奶品质有改善作用。%Selected 60 simmental hybrid dual-purpose cows under the principle of similar age, body weight, body condition at ifrst time parturition, and at about 200 days in lactation period. The cows were divided into 6 experimental groups which have 10 cows each. The basal diet composed of corn silage, alfalfa meal, wheat straw and concentrates. The experimental groups were given diets with different percentage of ensilaged apple pomace instead of same weight of corn silage (5kg、10kg、15kg、20kg、25kg). The results showed that, the milk production and milk quality indices, such as milk fat percentage, milk protein percentage of the ensilaged apple pomace added groups improved at different degree. Moreover, the group with 50% ensilaged apple pomace added improved significantly(P0.05)markedly. Ensilaged apple pomace could improve milking performance and milk quality of simmental hybrid dual-purpose cows at certain degree which varies with the amount of ensilaged apple pomace.

  9. 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.

  10. 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.

  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. 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.

  13. Double papilla flap technique for dual purpose

    Directory of Open Access Journals (Sweden)

    P Mohan Kumar

    2012-01-01

    Full Text Available Marginal tissue recession exposes the anatomic root on the teeth, which gives rise to -common patient complaints. It is associated with sensitivity, tissue irritation, cervical abrasions, and esthetic concerns. Various types of soft tissue grafts may be performed when recession is deep and marginal tissue health cannot be maintained. Double papilla flap is an alternative technique to cover isolated recessions and correct gingival defects in areas of insufficient attached gingiva, not suitable for a lateral sliding flap. This technique offers the advantages of dual blood supply and denudation of interdental bone only, which is less susceptible to permanent damage after surgical exposure. It also offers the advantage of quicker healing in the donor site and reduces the risk of facial bone height loss. This case report presents the advantages of double papilla flap in enhancing esthetic and functional outcome of the patient.

  14. A dual purpose Compton suppression spectrometer

    CERN Document Server

    Parus, J; Raab, W; Donohue, D

    2003-01-01

    A gamma-ray spectrometer with a passive and an active shield is described. It consists of a HPGe coaxial detector of 42% efficiency and 4 NaI(Tl) detectors. The energy output pulses of the Ge detector are delivered into the 3 spectrometry chains giving the normal, anti- and coincidence spectra. From the spectra of a number of sup 1 sup 3 sup 7 Cs and sup 6 sup 0 Co sources a Compton suppression factor, SF and a Compton reduction factor, RF, as the parameters characterizing the system performance, were calculated as a function of energy and source activity and compared with those given in literature. The natural background is reduced about 8 times in the anticoincidence mode of operation, compared to the normal spectrum which results in decreasing the detection limits for non-coincident gamma-rays up to a factor of 3. In the presence of other gamma-ray activities, in the range from 5 to 11 kBq, non- and coincident, the detection limits can be decreased for some nuclides by a factor of 3 to 5.7.

  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. Power and heat generation from wood wastes. At a dual-purpose power plant with a rotary tubular kiln; Aus Holzreststoffen Strom und Waerme gewinnen. Im Heizkraftwerk mit Drehrohrofenfeuerung

    Energy Technology Data Exchange (ETDEWEB)

    Hofstetter, E.M.; Letz, G. [Heinz H. Lorenz KG, Geretsried (Germany)

    1997-06-01

    Wood and related fuels tend to vary considerably as to calorific value, moisture content and particle size and burn with different efficiency. Therefore, new combustion-technical solution concepts are required that meet growing demands on combustion quality and automatic control response in the case of different fuel properties. The company Heinz H. Lorenz at Geretsried are pioneering a wood-fired dual-purpose power plant supplying energy and disposing of wood wastes. The project is sponsored by the Federal Environment Agency and the Bavarian ministry of economy (orig.) [Deutsch] Holz und artverwandte Brennstoffe schwanken in der Praxis oft erheblich in ihrem Heizwert, Feuchtegehalt, Stueckgroesse und verbrennen daher auch unterschiedlich gut. Gefragt sich deshalb neue Loesungen in der Feuerungstechnik angesichts steigender Anforderungen an Verbrennungsqualitaet und Regelverhalten bei unterschiedlichen Brennstoffeigenschaften. Neue Wege bei der Erstellung einer holzbefeuerten Heizkraftwerksanlage zur innerbetrieblichen Energieversorgung und Restholzentsorgung hat die Fa. Heinz H. Lorenz KG in Geretsried beschritten. Unterstuetzt wurde sie dabei durch das Umweltbundesamt und das bayerische Wirtschaftsministerium. (orig.)

  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. 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)

  20. 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.

  1. 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)

  2. Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, R.J.

    1998-11-17

    This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE).

  3. Interactive Effect of GA3, N and P Ameliorate Growth, Seed and Fibre Yield by Enhancing Photosynthetic Capacity and Carbonic Anhydrase Activity of Linseed:A Dual Purpose Crop

    Institute of Scientific and Technical Information of China (English)

    Mohammad N Khan; Firoz Mohammad

    2013-01-01

    Linseed (Linum usitatissimum L.) is an important dual-purpose, industrial crop. Its seeds are used for the extraction of oil and stem for fibres. However, the production of linseed is not going parallel with the increasing demand of its products. The present work was carried out with an aim to find out whether exogenous application of gibberellic acid (GA3) with or without graded levels of nitrogen (N) and phosphorus (P) could improve the performance of three linseed genotypes Parvati, Shekhar and Shubhra together with minimizing the costly fertilizer input and losses. Four combinations of N and P, viz., 0 mg N+0 mg P kg-1 soil (N0P0), N13.4P4.46, N26.8P8.94 and N40.2P13.4 were constituted. Half dose of each combination was applied basally at the time of sowing and remaining half dose was given at 40 d after sowing (DAS) as foliar spray along with 10-6 mol L-1 GA3. Prior to sowing, the seeds of each linseed genotype were grouped in to two, one group of seeds was soaked in 0 mol L-1 GA3 (control) and the other group was soaked in 10-6 M GA3 solution, each for 8 hours. Treatments were comprised of (i) 0 mol L-1 GA3+N0P0 (T0, control), (ii) 10-6 mol L-1 GA3+N13.4P4.46 (T1), (iii) 10-6 mol L-1 GA3+N26.8P8.94 (T2) and (iv) 10-6 mol L-1 GA3+N40.2P13.4 (T3). The crop performance was assessed in terms of growth, physiological and biochemical parameters at 60 and 75 DAS and yield attributes at harvest (175 DAS). The results showed a parallel increase in most of the parameters with increasing levels of N and P. However, application of 10-6 mol L-1 GA3 in association with N26.8P8.94 proved best, it enhanced seed yield, oil yield and fibre yield by 83.3, 97.3 and 78.7%, respectively accompanied with increase in net photosynthetic rate, carbonic anhydrase activity and dry matter accumulation. Among the genotypes tested, Shubhra performed best, while Parvati the least for most of the parameters studied. Thus, combined application of 10-6 mol L-1 GA3 plus N26.8P8.94 proved best

  4. 风光互补型发电和提水两用机组的研究%Research on wind and solar powers mutually complemented generating-pumping dual purpose unit

    Institute of Scientific and Technical Information of China (English)

    孙彦君; 尹钢吉; 滕云; 司振江; 陶延怀

    2012-01-01

    In accordance with the problems from high energy consumption and large pollution emission over the relevant national standards during pumping water with the conventional energies generally existed in China, a new generation of wind and solar powers mutually complemented generating-pumping dual purpose unit is developed with the key techniques, I. E. High-efficiency centrifugal pump, solar cell array, Darrieus vertical-axis wind turbine(VAWT) and "S" type low wind speed windmill, etc. , so as to largely save fuels and electric power resources and lower the production costs concerned. With the optimal structure, this kind of unit has realized the diversification of both the function and the application, and then got the target of energy saving and consumption reducing along with the enhancement of production efficiency.%针对我国目前普遍存在的利用常规能源来提水时能耗大和污染排放超过国家标准等问题,创制出新一代风光互补型发电和提水两用机组,采用高效离心潜水电泵、太阳能电池组和达里厄型立轴式风力机、S型低风速启动风车等这些关键技术来节约大量燃料、电能资源,降低生产成本.这种机组以最优的结构匹配,实现了功能和用途的多样化,达到了节能降耗及提高生产效率之目的.

  5. Breeding of a New Dual-Purpose Hot Pepper Variety — Jiahe 1%干鲜两用型辣椒新品种嘉禾1号的选育

    Institute of Scientific and Technical Information of China (English)

    庞凤琴; 鲍岩峰; 李玉珍; 于晓祥; 秦艳玲

    2011-01-01

    嘉禾1号是以国外品种经多代分离纯化选育出的自交系7310102为母本,自制杂交种经多代分离纯化选育出的自交系818481为父本杂交而成的干鲜两用型辣椒一代杂种。其植株生长势强,商品性好,抗逆性强,比当地主推品种增产36.8%。果实圆筒型,顶部钝尖,青熟果深绿而有光泽,红熟果油亮紫红色,果长13~15cm,果肩宽2.2~2.5cm,鲜椒皮厚3mm,干椒皮厚0.4mm。%Jiahe 1 is a hot pepper F1 hybrid of Inbred Line 7310102 from a foreign variety ×Inbred Line 818481 from a local hylrid variety, and a dual-purpose variety for fresh and dry fruit production. It is of strong growth vigour, high resistance, and the yield potential 36.8% higher than local main varieties. It grows glossy cylinder-shaped fruit with blunt tip from dark-green to purple-red, 13~15cm in length,2.2~2.5 cm in diameter, 3 mm in fresh fruit wall thichness and 0.4 mm in dry fruit wall thickness, with good marketability.

  6. Oxidation of nuclear fuel below 400 deg. Consequence on long-term dry storage; L'oxydation du combustible nucleaire au-dessous de 400 deg. Consequences sur l'entreposage a sec de longue duree

    Energy Technology Data Exchange (ETDEWEB)

    Dehaudt, Ph

    2000-07-01

    This document reviews the status of the knowledge on the oxidation of fuels below 400 deg C, in all its forms, including fuel rods, by examining the consequences of this reaction on the strength or ruin of the fuel rods during dry storage in air for a hundred years. The data available in the scientific literature, and the data acquired by CEA, are abundant on irradiated powders and pellets, but sparser for irradiated fuel fragments and for rods or sections of fuel rods. A bibliographic review is made to identify the morphological and structural changes, as well as the kinetic laws. An analysis and a summary is made with a concern to evaluate the risks of rod ruin by oxidation. The final section, in a few pages, addresses the essential lessons from this study. It presents: first, a summary of the main results of this review and its analysis, recommendations and remedies for storage; proposed research guidelines as well as precise topics, in order to fill out our knowledge and, even better, to identify the acceptable limits for storage. (author)

  7. Preservation of H2 production activity in nanoporous latex coatings of Rhodopseudomonas palustris CGA009 during dry storage at ambient temperatures: Preservation of R. palustris latex coatings

    Energy Technology Data Exchange (ETDEWEB)

    Piskorska, M. [Univ. of South Carolina, Aiken, SC (United States); Soule, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gosse, J. L. [North Carolina State Univ., Raleigh, NC (United States); Milliken, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flickinger, M. C. [North Carolina State Univ., Raleigh, NC (United States); Smith, G. W. [Univ. of South Carolina, Aiken, SC (United States); Yeager, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-01-21

    To assess the applicability of latex cell coatings as an ‘off-the-shelf’ biocatalyst, the effect of osmoprotectants, temperature, humidity and O2 on preservation of H2 production in Rhodopseudomonas palustris coatings was evaluated. Immediately following latex coating coalescence (24 h) and for up to 2 weeks of dry storage, rehydrated coatings containing different osmoprotectants displayed similar rates of H2 production. Beyond 2 weeks of storage, sorbitol-treated coatings lost all H2 production activity, whereas considerable H2 production was still detected in sucrose- and trehalose-stabilized coatings. The relative humidity level at which the coatings were stored had a significant impact on the recovery and subsequent rates of H2 production. After 4 weeks storage under air at 60% humidity, coatings produced only trace amounts of H2 (0–0.1% headspace accumulation), whereas those stored at < 5% humidity retained 27–53% of their H2 production activity after 8 weeks of storage. In conWhen stored in argon at < 5% humidity and room temperature, R. palustris coatings retained full H2 production activity for 3 months, implicating oxidative damage as a key factor limiting coating storage. Overall, the results demonstrate that biocatalytic latex coatings are an attractive cell immobilization platform for preservation of bioactivity in the dry state.

  8. Shielding calculations with SCALE/MAVRIC and comparison with measurements for the TN85 cask with vitrified high level radioactive waste

    Science.gov (United States)

    Thiele, Holger; Börst, Frank-Michael

    2017-09-01

    A series of dose rate/spectra measurements in the German interim storage facility Gorleben was carried out at a TN85 cask in April 2009. This type of cask is used for the transport and interim storage of vitrified high level radioactive waste (HAW) from reprocessing. The aim of this work is to assess the shielding component MAVRIC of the SCALE code system with these measurements for the use in the German Bundesamt für Kerntechnische Entsorgungssicherheit (BfE).

  9. 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.

  10. 某型通用机枪大杠杆曲线参数优化研究%Parameter Optimization of Big-lever Curve of a Dual-purpose Machine Gun

    Institute of Scientific and Technical Information of China (English)

    王瑞林; 张本军; 李永建; 郝刚

    2012-01-01

    为解决某型通用机枪首发装填困难的故障,利用线位移传感器进行了首发装填试验,得到了影响首发装填的重要因素,即拨弹滑板在向外和向里2个阶段所受阻力传递到装填拉柄运动方向上的最大值;在大杠杆的两段曲线上各取5个点,利用三次样条曲线对这两条曲线进行拟合,求出了2个最大值与插值点之间的函数关系;分别以这2个最大值为目标函数,以各插值点的Y向坐标为待优化参数,利用遗传算法对大杠杆曲线进行了优化,使得2个目标函数分别减少了47.24%和39.3%,达到了优化的目的.为该枪的改进工作提供指导,所使用的方法为同类型武器的设计和改进提供参考.%To solve the problem in first-round loading of a dual-purpose machine gun, the loading test was carried out by using a linear displacement sensor, and the importance factors which affect the first-round loading were revealed, i. e. two maximum values of resistances transferred from feed slide to the loading handle. Five points were taken in two segments of big-lever curve respectively, and the relationship between two maximum values and interpolation points were found out by using cubic spline curve fitting. Taken these two values as objective functions and the Y coordinates of interpolation points as the optimized parameters, the big-lever curves were optimized by using genetic algorithm, and the objective functions were reduced by 47. 24% and 39. 3% . The method can provide reference for the design and improvement of the similar weapons.

  11. LOAD-CHECK, program supported optimization of the fuel element disposal in cask CASTOR {sup registered} V casks; LOAD-CHECK, programmunterstuetzte Optimierung der Brennelemententsorgung in CASTOR {sup registered} V-Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Amian, D.; Braun, A. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Graf, R.; Hoffmann, V. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2010-05-15

    LOAD-CHECK is an interactive program module for the systematic and strategic spent fuel disposal planning. Using physical fuel element data the loading scenarios for the routine operation and the post-closure operation phase can be simulated for free selectable time periods. The basis for the loading license application are the available spent fuel casks according to the regulations of the interim storage facility. LOAD-CHECK allows the optimization of the loading campaigns with respect to the time schedule and the number of casks including the planning of optimized disposal of special spent fuel (MOX fuel elements or high-burnup fuel elements). Possibilities for a reduced post-closure operating phase of nuclear power plants might be the consequence.

  12. Delayed replantation after prolonged dry storage

    Directory of Open Access Journals (Sweden)

    Anita Rao

    2014-01-01

    Full Text Available Management of tooth avulsion in the permanent dentition often presents a challenge. Definitive treatment planning and consultation with specialists is seldom possible at the time of emergency treatment. Replantation of the avulsed tooth can restore esthetic appearance and occlusal function shortly after the injury. This article describes the management of a patient with an avulsed maxillary permanent incisor that had been air-dried for about 40 h. The replanted incisor retained its esthetic appearance and functionality 1 year after replantation, yet the long-term prognosis is not good because of progressive replacement root resorption.

  13. Efecto del régimen de defoliación sobre la producción de grano en trigo doble propósito Effect of the defoliation regime on grain production in dual purpose wheat

    Directory of Open Access Journals (Sweden)

    N. Peralta

    2011-06-01

    Full Text Available La utilización de cultivos de trigo con doble propósito contribuiría a realizar un uso más eficiente de los recursos ambientales. En Balcarce (Buenos Aires, Argentina se estableció un cultivo de trigo (Triticum aestivum, cv. Baguette 20 sembrado en fecha temprana (17 de marzo de 2006 para comparar la producción de forraje obtenida con tres regímenes de defoliación: T1 (tres defoliaciones, T2 (dos defoliaciones y TG (sin defoliaciones. En esos tres tratamientos y en otro no defoliado sembrado en fecha normal para la producción de grano (7 de junio, TN se cuantificó la biomasa aérea (g MS m², el rendimiento de grano (kg ha-1 y sus componentes. La producción de forraje de T1 fue similar que la de T2 (497 y 392 g m-2, respectivamente. Las defoliaciones posteriores al estado de dos nudos (estado Zadoks 32 afectaron negativamente la radiación fotosintéticamente activa interceptada por el cultivo, disminuyendo tanto el número de granos m-2 como el peso por grano; en consecuencia, la producción de grano cayó 60% respecto de TG y 71% respecto de TN. Se estableció una relación cuantitativa entre la disminución del rendimiento en grano y el momento del último corte (i.e., días entre la última defoliación y antesis. El trigo presenta la ventaja de poseer un destino alternativo a su uso forrajero, que no la presentan otros verdeos de invierno. Puede destinarse sólo para uso forrajero, sólo para producir grano, o se lo puede utilizar buscando un balance entre la producción de forraje y grano.The utilization of dual purpose wheat crops may contribute to perform a more efficient use of environmental resources. In Balcarce (Buenos Aires, Argentina a wheat crop (Triticum aestivum, cv. Baguette 20 was seeded at an early date (March 17, 2006 and the forage production was compared with three defoliation frequencies: T1 (three defoliations; T2 (two defoliations and TG (without defoliations. The aerial biomass, grain yield (kg ha-1 and its

  14. Niveles de infestación parasitaria y condición corporal en bovinos doble propósito infestados en condiciones naturales (Degree of infestation and body condition in dual purpose naturally infected cattle

    Directory of Open Access Journals (Sweden)

    Gustavo Morales

    2006-04-01

    work was carried out in a dual purpose farm oriented to milk production, located in San Antonio, Municipality of Pedraza in Barinas State. 65 cattle were coprologically examined by mean of McMaster technique with a oversaturated solution of NaCl as flotation liquid. The body condition of the animals was evaluated by mean the method of the scale with scores from 1 to 5 and 2,5 as inflection point. The packed cell volume was evaluated by mean of the microhaematocrit microcentrifugation method. The conjuntive color of each animal was evaluated as well. The highest e.p.g. count were observed in animals with a body condition ≤ 2,5. This animals can be considered as wormy animals, a small fraction of the herd showed a good body condition (≥ 3 and high e.p.g. count. This animals can be considered as resilient. Both group of animals are strong grass contaminants. The haematocrit values were similar between no wormy animals and resilient and higher than wormy animals. No association was found between haematocrit and conjuntive color. No anthelmintic resistance was diagnosed against the drugs used in the farm (Doramectine, Ivermectine and Albendazole.

  15. Costeño-201: nueva variedad de sorgo de temporal de doble propósito para Sinaloa Costeño-201: a new variety of rianfed sorghum of dual purpose in Sinaloa

    Directory of Open Access Journals (Sweden)

    Luis Alberto Hernández Espinal

    2011-10-01

    to commercial hybrids from private companies, which are grown in the region under the same conditions. Costeño-201 has better bromatological quality than commercial hybrids regarding to forage, with 9.5% protein and 64% digestibility, on average it exceeds in 2.7% and 5% respectively to commercial hybrids. It is tolerant to diseases that occur in the region, such as: ergot (Claviceps african, anthracnose (Colletotrichum graminicola, panicle blight (Fusarium moniliforme and charcoal stalk rot (Macrophomina phaseolina. Forage of sorghum Costeño-201, is recommended as a dual-purpose material in fodder conservation practices such as hay and silage.

  16. The application of fracture mechanics to the safety assessment of transport casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    BAM is the German responsible authority for the mechanical and thermal design safety assessment of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety proof of package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new ''Guidelines for the Application of Ductile Cast Iron for Transport and Storage Casks for Radioactive Materials''. Based on these guidelines higher stresses than before can become permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof with the help of the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA Advisory Material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will be concluded.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  3. Study of casks shielded with heavy metal to transport highly radioactive substances; Estudo de embalados com blindagem em metal pesado para transporte de substancias altamente radioativas

    Energy Technology Data Exchange (ETDEWEB)

    Lucchesi, R.F.; Hara, D.H.S.; Martinez, L.G.; Mucsi, C.S.; Rossi, J.L., E-mail: rflguimaraes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2014-07-01

    Nowadays, Brazil relies on casks produced abroad for transportation in its territory of substances that are sources of high radioactivity, especially the Mo-99. The product of the radioactive decay of the Mo-99 is the Tc-99m, which is used in nuclear medicine for administration to humans in the form of injectable radioactive drugs for the image diagnosis of numerous pathologies. This paper aims to study the existing casks in order to propose materials for the construction of the core part as shielding against gamma radiation. To this purpose, the existing literature on the subject was studied, as well as evaluation of existing and available casks. The study was focused on the core of which is made of heavy metals, especially depleted uranium for shielding the emitted radiation. (author)

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. Effects of T-type Channel on Natural Convection Flows in Airflow-Path of Concrete Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Gyeong Uk; Kim, Hyoung Jin; Cho, Chun Hyung [KORAD, Daejeon (Korea, Republic of)

    2016-05-15

    The natural convection flows occurring in airflow-path are not simple due to complex flow-path configurations such as horizontal ducts, bent tube and annular flow-path. In addition, 16 T type channels acting as the shroud are attached vertically and 16 channel supporting the canister are attached horizontally on the inner surface of over-pack. The existence and nonexistence of T type channels have influences on the flow fields in airflow- path. The concrete storage cask has to satisfy the requirements to secure the thermal integrity under the normal, off-normal, and accident conditions. The present work is aiming at investigating the effects of T type channels on the flows in airflow-path under the normal conditions using the FLUENT 16.1 code. In order to focus on the flows in airflow-path, fuel regions in the canister are regarded as a single cylinder with heat sources and other components are fully modeled. This study investigated the flow fields in airflow-path of concrete storage cask, numerically. It was found that excepting for the fuel regions, maximum temperatures on other components were evaluated below allowable values. The location of maximum velocities depended on support channels, T type channels and flow area. The flows through air inlets developed along annular flow- path with forming the hot plumes. According to the existence and nonexistence of T type channel, the plume behavior showed the different flow patterns.

  13. 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.

  14. 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

  15. 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

  16. Radiological characterization of container materials after interim storage using the example of CASTOR {sup registered} casks; Radiologische Charakterisierung von Behaelterwerkstoffen nach der Zwischenlagerung am Beispiel von CASTOR {sup registered} -Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Rother, W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Graf, R.; Winterhagen, D. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2012-11-01

    During interim storage the spent fuel casks or the casks for high-level waste (HAW) coquilles are exposed to neutron irradiation from their inventory that induces activation of the pannier the container body and its components. The presented contribution is an exemplary calculation of mean radial and axial reaction rates for components of HAW coquille or PWR fuel element loaded casks of the type CASTOR {sup registered} HAW28M and CASTOR {sup registered} V/19. The results of the Monte Carlo calculations include statistical uncertainties dependent on several parameters. Using the typical assumption that the casks will be unloaded after 40 years of interim storage the specific activities of significant reaction products are radiological assessed for further 60 years. An unrestricted release according paragraph 29 StrlSchV is discussed.

  17. 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

  18. 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.

  19. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  20. Summary of the technical review of the safety analysis reports for packaging (SARP) for the transnuclear transport/storage casks: TN-BRP and TN-REG

    Energy Technology Data Exchange (ETDEWEB)

    1986-07-01

    The Safety Analysis Reports for Packaging for two spent fuel shipping casks were technically reviewed by the Oak Ridge National Laboratory. The casks were designed by Transnuclear, Inc., for shipment of 85 Big Rock Point boiling water reactor fuel elements and 40 R.E. Ginna pressurized water reactor fuel elements from West Valley, New York, to Idaho Falls, Idaho. The intent of the review was to ensure compliance of the casks with the requirements the applicable Federal Regulations contained in 10 CFR Pt. 71 and allow issuance of Department of Energy Certificates of Compliance for transport by the Department of Energy Idaho Operations Office. The review was performed by a team of Oak Ridge National Laboratory staff assembled for their expertise in criticality analysis, shielding, metallurgy, nondestructive testing, thermal analysis, structural analysis, and containment. This report describes the review processes, the findings in each technical area, and the overall conclusion that a Certificate of Compliance could be issued for the proposed single shipment under the specified conditions and constraints.

  1. 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.

  2. 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.

  3. 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.

  4. Generation mean analysis of dual purpose traits in cowpea (Vigna ...

    African Journals Online (AJOL)

    Fadiji

    2012-06-07

    Jun 7, 2012 ... Key words: Gene effects, fodder, Vigna unguiculata, generation mean analysis. .... distribution pattern of fodder yield in the various crosses, the field ...... Tropical Agriculture (IITA), and Japan International Research Center.

  5. Dual-Purpose Bone Grafts Improve Healing and Reduce Infection

    Science.gov (United States)

    2011-08-01

    and PUR+BMP (High) (n = 14) groups. Two-thirds of the animals were used to assess bone formation, and the remainder was used to assess bacterial ...stabilized segmental defect in the rat femur. J Orthopaedic Res. 2005;23:816–823. 14. Anglen JO. Comparison of soap and antibiotic solutions for irrigation of

  6. ECONOMIC SUSTAINABILITY OF THE LOCAL DUAL-PURPOSE CATTLE

    Directory of Open Access Journals (Sweden)

    Zuzana Krupová

    2015-09-01

    Full Text Available Base economic characteristics (total revenues, total costs, profit and profitability ratio of the Slovak Pinzgau breed were calculated in this study. Under the actual production and economic conditions of the breed, production system is operated with loss (-457 € per cow and per year and with negative profitability ratio (-20%. Optimisation of the production parameters on the level defined in the breed standard (5,200 kg milk per cow and year, 92% for conception rate of cows, 404 days of calving interval and 550 g in daily gain of reared heifers and improved udder health traits (clinical mastitis incidence and somatic cells score was of positive impact on the total revenues (+34%, on the effective utilisation of costs (+105% and balanced profit of dairy systems. Next to the positive profitability of the system, higher quality and security of dairy milk products should be mentioned there. Moreover, direct subsidies as an important factor of positive economic result of dairy cattle systems has to be pointed as well. Subsidies should be provided to compensate the real biological limitation of the local breed farmed in marginal areas. However, improvement of the production parameters of the Slovak Pinzgau breed is recommended with the same attention to reach the economic sustainability of dairy production system. To reach economic sustainability of the breed from practical point of view, the farmer activity should be aimed especially to the enhanced herd management.

  7. development of a dual purpose refrigeration system for domestic use

    African Journals Online (AJOL)

    user

    and Air conditioning Engineers. BSR: Board of Standards ... whose elements were encased in a singular unit as is the case in modern ... production, operation and maintenance. 2. ..... conditioning) and food storage (refrigeration). Table 4: ...

  8. The Milking Performance of dual-purpose Crossbred Yaks

    Institute of Scientific and Technical Information of China (English)

    B.Jialin; W.Mingqiang; L.Zhonglin; J.M.Chesworth

    2005-01-01

    Wild yak males were crossed with domesticated female yaks to produce a breeding herd of crossbred animals. The milk production of the progeny of this herd was measured in the present study. Yaks were milked once daily after prior suckling of the calf. yaks were allowed to graze for 15 h/day on high-altitude(3200 to 3500 m) pastures. No supplementary food was given. Average daily milk yield of 1.77(s. e. 0. 16)kg and milk yield over 120 days,212.2(s. e. 20.2)kg were significantly higher(P<0.01)than the corresponding figures for uninlproved domestic yaks(1.53(s. e. 0. 10)kg and 164. 6(s. e. 10.5)kg respectively). The fat content of milk from crossbred yaks, 52(s. e. 2. 9)g/kg, was not significantly dlfferent from that of unimproved animals, 53.5(s. e. 4. 1), g/kg.

  9. Dual-Purpose Millikan Experiment with Polystyrene Spheres

    Science.gov (United States)

    Wall, C. N.; Christensen, F. E.

    1975-01-01

    This procedure, using polystyrene spheres of specified diameter, renders the Millikan oil drop experiment more accurate than the conventional procedure of the polystyrene spheres, eliminates size estimation error, and removes the guesswork involved in assigning proper index integers to the observed charges. (MLH)

  10. Specification of requirements to get a license for an Independent Spent Fuel Dry Storage Installation (ISFSI) at the site of the NPP-LV; Especificacion de los requerimientos para tramitar una licencia de una instalacion independiente de almacenamiento temporal en seco de combustible gastado (ISFSI) en el sitio de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Serrano R, M. L., E-mail: mlserrano@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    This article describes some of the work done in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) to define specifically the requirements that the Federal Electricity Commission (CFE) shall meet to submit for consideration of CNSNS an operation request of an Independent Spent Fuel Dry Storage Installation (ISFSI). The project of a facility of this type arose from the need to provide storage capacity for spent nuclear fuel in the nuclear power plant of Laguna Verde (NPP-LV) and to continue the operation at the same facility in a safe manner. The licensing of these facilities in the United States of America has two modes: specific license or general license. The characteristics of these licenses are described in this article. However, in Mexico the existing national legislation is not designed for such license types, in fact there is a lack of standards or regulations in this regard. The regulatory law of Article 27 of the Constitution in the nuclear matter, only generally establishes that this type of facility requires an authorization from the Ministry of Energy. For this reason and because there is not a national legislation, was necessary to use the legislation that provides the Nuclear Regulatory Commission of USA, the US NRC. However, it cannot be applied as is established, so was necessary that the CNSNS analyze one by one the requirements of both types of license and determine what would be required to NPP-LV to submit its operating license of ISFSI. The American regulatory applicable to an ISFSI, the 10-Cfr-72 of the US NRC, establishes the requirements for both types of licenses. Chapter 10-Cfr was analyzed in all its clauses and coupled to the laws, regulations and standards as well as to the requirements established by CNSNS, all associated with a store spent fuel on site; the respective certification of containers for spent fuel dry storage was not included in this article, even though the CNSNS also performed that activity under the

  11. SCALE/MAVRIC calculation of dose rates measured for a gamma radiation source in a thick-walled transport and storage cask of ductile cast iron with lead inserts

    Science.gov (United States)

    Baumgarten, Werner; Thiele, Holger; Ruprecht, Benjamin; Phlippen, Peter-W.; Schlömer, Luc

    2017-09-01

    Dose rate calculations are important for judging the shielding performance of transport casks for radioactive material. Therefore it is important to have reliable calculation tools. We report on measured and calculated dose rates near a thick-walled transport and storage cask of ductile cast iron with lead inserts and a Co-60 source inside. In a series of experiments the thickness of the inserts was varied, and measured dose rates near the cask were compared with SCALE/MAVRIC 6.1.3 and SCALE/MAVRIC 6.2 calculation results. Deviations from the measurements were found to be higher for increased lead thicknesses. Furthermore, it is shown how the shielding material density, air scattering and accounting for the floor influence the quality of the calculation.

  12. Study and full-scale test of a high-velocity grade-crossing simulated accident of a locomotive and a nuclear-spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Huerta, M.; Yoshimura, H.R.

    1983-02-01

    This report described structural analyses of a high-speed impact between a locomotive and a tractor-trailer system carrying a nuclear-spent-fuel shipping cask. The analyses included both mathematical and physical scale-modeling of the system. The report then describes the full-scale test conducted as part of the program. The system response is described in detail, and a comparison is made between the analyses and the actual hardware response as observed in the full-scale test. 34 figures.

  13. 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

  14. 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)

  15. Choice of optimum size of installations for dual-purpose production of desalted water and electricity, using nuclear power; Le dimensionnement optimum des installations de production mixte d'eau sessalee et d'electricite faisant intervenir l'energie nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J. [Commissariat a l' Energie Atomique, Paris (France)

    1966-07-01

    The author used a method starting with water and power demand curves; this leads to the rational allocation of production costs to water and power within a given market. The power demand curve is needed as it seems improbable to sell at a constant price the enormous quantity of electricity produced by a dual purpose plant. Criteria based on principles of classical economics, help to select objectively desalination methods and plant sizes. On these criteria, normative methods for tariffing action of water and power can be based, while adhering as closely as possible to structure of demand. Examples of such criteria are the maximum profit of the supplier or the maximum satisfaction of the consumers taken collectively. In the first case marginal costs must be equated to marginal revenue, in the second one marginal cost to marginal satisfaction (theory of surpluses). The plant size often determines the choice of desalination process. Therefore the shape of the water demand curve and the economic criterion adopted (public or private ownership, capital restrictions etc.) often determine in this way both size and type of plant. Before deciding on the desalination technique, market surveys and rather subtle economic analyses are therefore necessary. (author) [French] Le probleme est presente en introduisant la notion de courbes de demande d'eau et d'electricite, ce qui permet d'aboutir a un partage rationnel des couts de revient entre eau et electricite dans ]e cadre d'un marche. L'objet de l'etude est, a partir des principes de l'economie classique, de donner des criteres objectifs de selection des dimensions des installations et des techniques de dessalement et d'en deduire une methode normative de tarification des deux produits lies: eau et electricite, en collant autant que possible a la structure de la demande. Ces criteres sont en particulier, soit le maximum de benefice de l'exploitant, soit le maximum de satisfaction des

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  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. 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.)

  2. 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).

  3. 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

  4. 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

  5. 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.

  6. Safety aspects of dry spent fuel storage and spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Botsch, Wolfgang; Smalian, Silva; Hinterding, Peter [TUV NORD EnSys Hannover, GmbH and Co. KG, Hanover (Germany); Volzke, Holger; Wolff, Dietmar; Kasparek, Eva-Maria [BAM Federal Institute for Materials Research and Testing, Berlin (Germany)

    2013-07-01

    As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TUV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  7. Dual purpose use of preterm piglets as a model of pediatric GI disease

    Science.gov (United States)

    Necrotizing enterocolitis (NEC) is the most common gastrointestinal complication in human neonates, yet the pathogenesis of this disease remains poorly understood. A fundamental approach to understanding the etiology and underlying biology of NEC is the use of in vivo experimental animal models, pri...

  8. Factor analysis for genetic evaluation of linear type traits in dual purpose breeds

    Directory of Open Access Journals (Sweden)

    B. Contiero

    2010-01-01

    Full Text Available Type traits have become a selection goal in many breeder association due to the economic value recognised to some measurements. In dairy cattle, relationships of type with production (Veerkamp and Brotherstone, 1997; Kadarmideen and Wegmann, 2003 and/or with longevity (Larroque and Ducrocq, 2001; Vukasinovic et al., 2002 have been widely investigated in order to use type traits as predictors of herd life.

  9. Dual purpose use of preterm piglets as a model of pediatric GI disease

    DEFF Research Database (Denmark)

    Oosterloo, Berthe C; Premkumar, Muralidhar; Stoll, Barbara

    2014-01-01

    Necrotizing enterocolitis (NEC) is the most common gastrointestinal complication in human neonates, yet the pathogenesis of this disease remains poorly understood. A fundamental approach to understanding the etiology and underlying biology of NEC is the use of in vivo experimental animal models, ...

  10. Property of Dual-purposed Maca Powder as Medicine and Food

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    [Objective] The aim was to research property of maca and three plants' powders and process parameters of compound maca direct compression. [Method] Based on analysis data from SAS, bulk density, tap density, angle of repose and swelling of powders were studied as per single factor method and orthogonal exper- imental design. [Result] The test indicated that fillibilities of plant A and B, and maca powders are better and flowability plays an important role in fractional close of compound maca powder; plant A and B powders have a significant effect on bulk density of maca (P=0.0125), an extremely significant effect on swelling volume ratio (P=0.008 9) and little effect on tap density (g/ml); the optimal process condition of compound maca powder is as follows: A at 0.15 share; B at 0.10 share; C at 0.05 share; the optimal swelling volume ratio is at 2.459. [Conclusion] The technology is reasonable in formulation and satisfactory in fillibility, swelling ability, flowability, and it could serve as theoretical basis for the industrial production of maca tablets.

  11. A dual purpose universal influenza vaccine candidate confers protective immunity against anthrax.

    Science.gov (United States)

    Arévalo, Maria T; Li, Junwei; Diaz-Arévalo, Diana; Chen, Yanping; Navarro, Ashley; Wu, Lihong; Yan, Yongyong; Zeng, Mingtao

    2017-03-01

    Preventive influenza vaccines must be reformulated annually because of antigen shift and drift of circulating influenza viral strains. However, seasonal vaccines do not always match the circulating strains, and there is the ever-present threat that avian influenza viruses may adapt to humans. Hence, a universal influenza vaccine is needed to provide protective immunity against a broad range of influenza viruses. We designed an influenza antigen consisting of three tandem M2e repeats plus HA2, in combination with a detoxified anthrax oedema toxin delivery system (EFn plus PA) to enhance immune responses. The EFn-3×M2e-HA2 plus PA vaccine formulation elicited robust, antigen-specific, IgG responses; and was protective against heterologous influenza viral challenge when intranasally delivered to mice three times. Moreover, use of the detoxified anthrax toxin system as an adjuvant had the additional benefit of generating protective immunity against anthrax. Hence, this novel vaccine strategy could potentially address two major emerging public health and biodefence threats. © 2016 John Wiley & Sons Ltd.

  12. A ferrite LTCC based dual purpose helical antenna providing bias for tunability

    KAUST Repository

    Ghaffar, Farhan A.

    2015-03-30

    Typically, magnetically tunable antennas utilize large external magnets or coils to provide the magneto-static bias. In this work, we present a novel concept of combining the antenna and the bias coil in one structure. A helical antenna has been optimized to act as the bias coil in a ten layer ferrite LTCC package, thus performing two functions. This not only reduces the overall size of the system by getting rid of the external bias source but also eliminates demagnetization effect (fields lost at air-to-substrate interface), which reduces the required magneto-static field strength and makes the design efficient. RF choking inductor and DC blocking capacitor have been monolithically integrated as package elements to allow the magnetostatic and microwave excitation at the same time. The design has been optimized for its low frequency and high frequency performance in two different simulators. A measured tuning range of 10% is achieved at a center frequency of 13 GHz. The design is highly suitable for low cost, compact, light-weight and tunable microwave systems. © 2002-2011 IEEE.

  13. Dual Purpose Simulation: New Data Link Test and Performance Limit Testing of Currently Deployed Data Link

    Science.gov (United States)

    Robinson, Daryl C.

    2002-01-01

    While the results of this paper are similar to those of [I], in this paper technical difficulties present in [I] are eliminated, producing better results, enabling one to more readily see the benefits of Prioritized CSMA (PCSMA). A new analysis section also helps to generalize this research so that it is not limited to exploration of the new concept of PCSMA. Commercially available network simulation software, OPNET version 7.0, simulations are presented involving an important application of the Aeronautical Telecommunications Network (ATN), Controller Pilot Data Link Communications (CPDLC) over the Very High Frequency Data Link Mode 2 (VDL-2). Communication is modeled for essentially all incoming and outgoing nonstop air-traffic for just three United States cities: Cleveland, Cincinnati, and Detroit. The simulation involves 111 Air Traffic Control (ATC) ground stations, 32 airports distributed throughout the U.S., which are either sources or destinations for the air traffic landing or departing from the three cities, and also 1,235 equally equipped aircraft-taking off, flying realistic free-flight trajectories, and landing in a 24-hr period. Collision-less PCSMA is successfully tested and compared with the traditional CSMA typically associated with VDL-2. The performance measures include latency, throughput, and packet loss. As expected, PCSMA is much quicker and more efficient than traditional CSMA. These simulation results show the potency of PCSMA for implementing low latency, high throughput and efficient connectivity. Moreover, since PCSMA outperforms traditional CSMA, by simulating with it, we can determine the limits of performance beyond which traditional CSMA may not pass. So we have the tools to determine the traffic-loading conditions where traditional CSMA will fail, and we are testing a new and better data link that could replace it with relative ease. Work is currently being done to drastically expand the number of flights to make the simulation more representative of the National Aerospace System.

  14. Targeting pharmacoresistant epilepsy and epileptogenesis with a dual-purpose antiepileptic drug.

    Science.gov (United States)

    Doeser, Anna; Dickhof, Gesa; Reitze, Margit; Uebachs, Mischa; Schaub, Christina; Pires, Nuno Miguel; Bonifácio, Maria João; Soares-da-Silva, Patrício; Beck, Heinz

    2015-02-01

    In human epilepsy, pharmacoresistance to antiepileptic drug therapy is a major problem affecting a substantial fraction of patients. Many of the currently available antiepileptic drugs target voltage-gated sodium channels, leading to a rate-dependent suppression of neuronal discharge. A loss of use-dependent block has emerged as a potential cellular mechanism of pharmacoresistance for anticonvulsants acting on voltage-gated sodium channels. There is a need both for compounds that overcome this resistance mechanism and for novel drugs that inhibit the process of epileptogenesis. We show that eslicarbazepine acetate, a once-daily antiepileptic drug, may constitute a candidate compound that addresses both issues. Eslicarbazepine acetate is converted extensively to eslicarbazepine after oral administration. We have first tested using patch-clamp recording in human and rat hippocampal slices if eslicarbazepine, the major active metabolite of eslicarbazepine acetate, shows maintained activity in chronically epileptic tissue. We show that eslicarbazepine exhibits maintained use-dependent blocking effects both in human and experimental epilepsy with significant add-on effects to carbamazepine in human epilepsy. Second, we show that eslicarbazepine acetate also inhibits Cav3.2 T-type Ca(2+) channels, which have been shown to be key mediators of epileptogenesis. We then examined if transitory administration of eslicarbazepine acetate (once daily for 6 weeks, 150 mg/kg or 300 mg/kg) after induction of epilepsy in mice has an effect on the development of chronic seizures and neuropathological correlates of chronic epilepsy. We found that eslicarbazepine acetate exhibits strong antiepileptogenic effects in experimental epilepsy. EEG monitoring showed that transitory eslicarbazepine acetate treatment resulted in a significant decrease in seizure activity at the chronic state, 8 weeks after the end of treatment. Moreover, eslicarbazepine acetate treatment resulted in a significant decrease in mossy fibre sprouting into the inner molecular layer of pilocarpine-injected mice, as detected by Timm staining. In addition, epileptic animals treated with 150 mg/kg, but not those that received 300 mg/kg eslicarbazepine acetate showed an attenuated neuronal loss. These results indicate that eslicarbazepine potentially overcomes a cellular resistance mechanism to conventional antiepileptic drugs and at the same time constitutes a potent antiepileptogenic agent. © The Author (2014). Published by Oxford University Press on behalf of the Guarantors of Brain. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  15. Dual-purpose self-deliverable lunar surface PV electrical power system

    Science.gov (United States)

    Arnold, Jack H.; Harris, David W.; Cross, Eldon R.; Flood, Dennis J.

    1991-01-01

    A safe haven and work supported PV power systems on the lunar surface will likely be required by NASA in support of the manned outpost scheduled for the post-2000 lunar/Mars exploration and colonization initiative. Initial system modeling and computer analysis shows that the concept is workable and contains no major high risk technology issues which cannot be resolved in the circa 2000 to 2025 timeframe. A specific selection of the best suited type of electric thruster has not been done; the initial modeling was done using an ion thruster, but Rocketdyne must also evaluate arc and resisto-jets before a final design can be formulated. As a general observation, it appears that such a system can deliver itself to the Moon using many system elements that must be transported as dead payload mass in more conventional delivery modes. It further appears that a larger power system providing a much higher safe haven power level is feasible if this delivery system is implemented, perhaps even sufficient to permit resource prospecting and/or lab experimentation. The concept permits growth and can be expanded to include cargo transport such as habitat and working modules. In short, the combined payload could be manned soon after landing and checkout. NASA has expended substantial resources in the development of electric propulsion concepts and hardware that can be applied to a lunar transport system such as described herein. In short, the paper may represent a viable mission on which previous investments play an invaluable role. A more comprehensive technical paper which embodies second generation analysis and system size will be prepared for near-term presentation.

  16. Biophysical and economic water productivity of dual-purpose cattle farming.

    Science.gov (United States)

    Sraïri, M T; Benjelloun, R; Karrou, M; Ates, S; Kuper, M

    2016-02-01

    This study analyzes key factors influencing water productivity in cattle rearing, particularly in contexts characterized by water scarcity. This was done through year-round monitoring of on-farm practices within five smallholder farms located in the Saïss area (northern Morocco). The on-farm monitoring protocol consisted of characterizing: (i) volumes of water used for fodder production and distinguished by source (rainfall, surface irrigation and groundwater), (ii) virtual water contained in off-farm feed resources, (iii) total forage biomass production, (iv) dietary rations fed to lactating cows and their calves and (v) milk output and live weight gain. Findings reveal a mean water footprint of 1.62±0.81 and 8.44±1.09 m3/kg of milk and of live weight gain, respectively. Groundwater represented only 13.1% and 2.2% of the total water used to get milk and live weight gain, respectively, while rainfall represented 53.0% and 48.1% of the total water for milk and live weight gain, respectively. The remaining water volumes used came from surface irrigation water (7.4% for milk and 4.0% for live weight gain) and from virtual water (26.5% for milk and 44.7% for live weight gain). The results also revealed a relatively small gross margin per m3 of water used by the herd, not exceeding an average value of US $ 0.05, when considering both milk and live weight. Given the large variability in farm performances, which affect water productivity in cattle rearing throughout the production process, we highlight the potential for introducing a series of interventions that are aimed at saving water, while concurrently improving efficiency in milk production and live weight gain. These interventions should target the chain of production functions that are implemented throughout the process of water productivity in cattle rearing. Moreover, these interventions are of particular importance given our findings that livestock production depends largely upon rainfall, rather than groundwater, in an area afflicted with sustained droughts, overexploitation of groundwater resources and growing water scarcity.

  17. Dual Purpose Landscaping Tools: Small Extra Dimensions in AdS/CFT

    Energy Technology Data Exchange (ETDEWEB)

    Polchinski, Joseph; /Santa Barbara, KITP /UC, Santa Barbara; Silverstein, Eva; /Santa Barbara, KITP /UC, Santa Barbara /SLAC /Stanford U., Phys. Dept.

    2010-08-26

    We propose a class of AdS/CFT dual pairs which have small internal dimensions on the gravity side. Starting from known Freund-Rubin AdS/CFT dual pairs, we use 7-branes to nearly cancel the curvature energy of the internal dimensions while maintaining their stabilization. This leads to a new corner of the landscape - a class of AdS solutions with a hierarchically large AdS radius - with a dual field theory given (implicitly) by the infrared limit of a concrete brane construction involving D3-branes, 7-branes, and curvature. We first construct a class of hierarchical AdS5/CFT4 dual pairs with a simple formula for the number of degrees of freedom which we interpret in the dual QFT. We then generalize these to AdS4/CFT3 duals, and suggest extensions of the method to obtain de Sitter solutions.

  18. Generator, mechanical, smoke: For dual-purpose unit, XM56, Yuma Proving Ground, Yuma, Arizona

    Energy Technology Data Exchange (ETDEWEB)

    Driver, C.J.; Ligotke, M.W.; Moore, E.B. Jr. (Pacific Northwest Lab., Richland, WA (United States)); Bowers, J.F. (Dugway Proving Ground, UT (United States))

    1991-10-01

    The US Army Chemical Research, Development and Engineering Center (CRDEC) is planning to perform a field test of the XM56 smoke generator at the US Army Yuma Proving Ground (YPG), Arizona. The XM56, enabling the use of fog oil in combination with other materials, such as graphite flakes, is part of an effort to improve the efficiency of smoke generation and to extend the effectiveness of the resulting obscurant cloud to include the infrared spectrum. The plan field operation includes a road test and concurrent smoke- generation trials. Three M1037 vehicles with operation XM56 generators will be road-tested for 100 h. Smoke will be generated for 30 min from a single stationary XM56 four times during the road test, resulting in a total of 120 min of smoke generation. The total aerial release of obscurant materials during this test is expected to be 556 kg (1,220 lb) of fog oil and 547 kg (1,200 lb) of graphite flakes. This environmental assessment has evaluated the consequences of the proposed action. Air concentrations and surface deposition levels were estimated using an atmospheric dispersion model. Degradation of fog oil and incorporation of graphite in the soil column will limit the residual impacts of the planned action. No significant impacts to air, water, and soil quality are anticipated. risks to the environment posed by the proposed action were determined to be minimal or below levels previously found to pose measurable impacts. Cultural resources are present on YPG and have been identified in adjacent areas; therefore, off-road activities should be preceded by a cultural resource survey. A Finding of No Significant Impact is recommended. 61 refs., 1 fig.

  19. 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.

  20. 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

  1. Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Theodore E. Madey

    2001-10-01

    Characterize the effects of temperature and radiation processes on the interactions of H20 with oxide surfaces. Our experiments focused on the fundamental interaction of H20 molecules with surfaces of U02. We characterized the surface chemistry of adsorbed H2O using thermal desorption methods and radiotracer methods, as well as x-ray photoelectron spectroscopy (XPS) and low energy ion scattering (LEIS). In parallel with these measurements of thermal effects, we examined the effects of secondary electrons and high-energy photons on hydrogen and oxygen generation and, and how this related to corrosion of spent nuclear fuel. These studies concentrated on neutral and ionic (cation and anion) desorption products of low-energy electron irradiation of water-covered UO2.

  2. Evaluation of Dried Storage of Platelets and RBC for Transfusion: Lyophilization and Other Dehydration Techniques

    Science.gov (United States)

    1991-01-31

    storage as a transfusion resource for combat casualty care. The work is being carried out at three collaborating medical centers: in Greenville, NC...optimization of stabilization procedures currently underway will likely verify a patentable preparation protocol. As an interinstitutional agreement is...conditions). These modified para-plts are being evaluated in the collaborating laboratories. Results are not yet consistent, but encouraging. Whether or not

  3. Replantation after extended dry storage of avulsed permanent incisors: report of a case.

    Science.gov (United States)

    Cobankara, Funda Kont; Ungor, Mete

    2007-08-01

    A 15-year-old boy lost his maxillary right and left central incisor teeth in a bicycle accident. He was referred to our clinic 1 week after the injury. The crown-root integrities of both the teeth were not damaged. Although the teeth were stored under dry conditions for 1 week, reimplantation of the teeth was planned to retain the teeth in the mouth for as long a period as possible because of the patient's age. Following the debridement and sterilization of root surfaces in 2.5% NaOCl, root canals were prepared and filled with calcium hydroxide. Then, about 2 mm of the apexes were resected to ensure that the roots easily seated in the alveolar socket and the prepared cavities in root ends were obturated with the amalgam. The teeth were placed into their respective sockets and splinted temporarily. The root canal therapy was completed 5 weeks later. Ankylosis was observed radiographically after 10 months. The patient is now 23 years old and he is still able to use both the central incisors functionally. However, there is a pink appearance on the cervical buccal surface of left central incisor because of progressive replacement resorption. In this case, the new treatment plan is to perform a permanent restoration with dental implants following the extraction of both teeth. Even though the long-term prognosis is uncertain, this treatment technique has provided an advantage for the patient in his adolescent period by maintaining the height of alveolar bone and making the provision of an aesthetically acceptable permanent restoration at a later age possible.

  4. 78 FR 19148 - Shielding and Radiation Protection Review Effort and Licensing Conditions for Dry Storage...

    Science.gov (United States)

    2013-03-29

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 Shielding and Radiation Protection Review Effort and...-ISG-26A), Revision 0, ``Shielding and Radiation Protection Review Effort and Licensing Conditions for... to NRC staff when reviewing the shielding and radiation protection portions of applications...

  5. Evaluation of maximum allowable temperature inside basket of dry storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Chae, Kyoung Myoung; Choi, Byung Il; Lee, Heung Young; Song, Myung Jae [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    This study provides a maximum allowable fuel temperature through a preliminary evaluation of the UO{sub 2} weight gain that may occur on a failed (breached sheathing) element of a fuel bundle. Intact bundles would not be affected as the UO{sub 2} would not be in contact with the air for the fuel storage basket. The analysis is made for the MACSTOR/KN-400 to be operated in Wolsong ambient air temperature conditions. The design basis fuel is a 6-year cooled fuel bundle that, on average has reached a burnup of 7,800 MWd/MTU. The fuel bundle considered for analysis is assumed to have a high burnup of 12,000 MWd/MTU and be located in a hot basket. The MACSTOR/KN-400 has the same air circuit as the MACSTOR and the air circuit will require a slightly higher temperature difference to exit the increased heat load. The maximum temperature of a high burnup bundle stored in the new MACSTOR/KN-400 is expected to be about 9 .deg. C higher than the fuel temperature of the MACSTOR at an equivalent constant ambient temperature. This temperature increase will in turn increase the UO{sub 2} weight gain from 0.06% (MACSTOR for Wolsong conditions) to an estimated 0.13% weight gain for the MACSTOR/KN-400. Compared to an acceptable UO{sub 2} weight gain of 0.6%, we are thus expecting to maintain a very acceptable safety factor of 4 to 5 for the new module against unacceptable stresses in the fuel sheathing. For the UO{sub 2} weight gain, the maximum allowable fuel temperature was shown by 164 .deg. C.

  6. CFD Analysis of Dry Storage System for CANDU Spent Fuel using Fluent 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Byung Soo; Jeong, Yong Hoon; Chang, Soon Heung Chang [KAIST, Taejon (Korea, Republic of)

    2006-07-01

    To obtain the licensing for MACSTOR/KN-400 developed by KHNP, 3-D CFD analysis was demanded to prove that the maximum temperature was not over the limited temperature (93 .deg. C locally and 66 .deg. C averagely for concrete). Though the thermal-hydraulic prediction by CATHENA-code show the reliable results, that could not the temperature distribution. That is, that could not predict the location of maximum temperature well. In this study, the analysis of the temperature distribution on the natural convection flow with thermal radiation shows the concrete temperature distribution. It was different from the predicted results by CATHENA-code. Therefore, to obtain the licensing for AMCSTOR/KN- 400, CFD analysis should be performed by 3-D CFD code like FLUENT at the same time.

  7. Application of nonlinear ultrasonics to inspection of stainless steel for dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Ulrich, Timothy James II [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderson, Brain E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Remillieux, Marcel C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Le Bas, Pierre -Yves [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pieczonka, Lukasz [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-22

    This report summarized technical work conducted by LANL staff an international collaborators in support of the UFD Storage Experimentation effort. The focus of the current technical work is on the detection and imaging of a failure mechanism known as stress corrosion cracking (SCC) in stainless steel using the nonlinear ultrasonic technique known as TREND. One of the difficulties faced in previous work is in finding samples that contain realistically sized SCC. This year such samples were obtained from EPRI. Reported here are measurements made on these samples. One of the key findings is the ability to detect subsurface changes to the direction in which a crack is penetrating into the sample. This result follows from last year's report that demonstrated the ability of TREND techniques to image features below the sample surface. A new collaboration was established with AGH University of Science and Technology, Krakow, Poland.

  8. 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

  9. 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)

  10. 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)

  11. 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。

  12. 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)

  13. 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.

  14. 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."

  15. 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.

  16. Conceptual design and cost study for a dual-purpose nuclear-electric reverse osmosis seawater conversion plant

    Energy Technology Data Exchange (ETDEWEB)

    1979-04-01

    The objective of this study was to develop a conceptual design and cost estimate for a 25 million gallon per day seawater reverse osmosis desalting plant operating at both Caribbean and Persian Gulf sites. The plant would operate in conjunction with a 1000 MW(e) nuclear power plant. Four seawater membrane manufacturers were supplied with feedwater analysis and a simplified cost estimating procedure in order to recommend membrane systems which would be applicable. For both sites a two-stage system was selected for development of a conceptual cost estimate. The product water cost was found to be (based upon 1978 United States construction costs) $3.17/1000 gallons for the Caribbean site and $3.75/1000 gallons for the Persian Gulf site.

  17. The Potentials for the Use of Single- versus Dual-Purpose Officers in Firms:

    DEFF Research Database (Denmark)

    Theotokas, Ioannis; Wagtmann, Maria Anne

    2010-01-01

    in maritime schools or universities. In our treatment of economic issues, we will focus on insights from the resource-based view of the firm, labour economics, and transaction cost economics, and then make some general statements about the potential economic advantages and disadvantages to a shipowner...

  18. Low-Resistance Dual-Purpose Air Filter Releasing Negative Ions and Effectively Capturing PM2.5.

    Science.gov (United States)

    Zhao, Xinglei; Li, Yuyao; Hua, Ting; Jiang, Pan; Yin, Xia; Yu, Jianyong; Ding, Bin

    2017-04-05

    The fatal danger of pollution due to particulate matter (PM) calls for both high-efficiency and low-resistance air purification materials, which also provide healthcare. This is however still a challenge. Herein, a low-resistance air filter capable of releasing negative ions (NIs) and efficiently capturing PM2.5 was prepared by electrospinning polyvinylidene fluoride (PVDF) fibers doped with negative ions powder (NIPs). The air-resistance of fibrous membranes decreased from 9.5 to 6 Pa (decrease of 36%) on decreasing the average fiber diameter from 1.16 to 0.41 μm. Moreover, the lower rising rate of air-resistance with reduction in pore size, for fibrous membranes with thinner fiber diameter was verified. In addition, a single PVDF/NIPs fiber was provided with strong surface potentials, due to high fluorine electronegativity, and tested using atomic force microscopy. This strong surface potential resulted in higher releasing amounts of NIs (RANIs). Interestingly, reduction of fiber diameter favored the alleviation of the shielding effects on electric field around fibers and promoted the RANIs from 798 to 1711 ions cc(-1). Moreover, by regulating the doping contents of NIPs, the RANIs increased from 1711 to 2818 ions cc(-1). The resultant fibrous membranes showed low air resistance of 40.5 Pa. Field-tests conducted in Shanghai showed stable PM2.5 purification efficiency of 99.99% at high RANIs, in the event of haze.

  19. THE DUAL PURPOSE OF TEACHING LITERATURE: TO PROVIDE STIMULATING COURSE CONTENT AND TO DEVELOP STUDENTS’ COMMUNICATIVE ABILITIES

    Directory of Open Access Journals (Sweden)

    Budiati Budiati

    2013-11-01

    Full Text Available It is the individual reader’s freedom to interpret a text according to his own outlook on the world that makes the study of literature such an exciting and liberating experience. This paper will look at some of the issues and ways in which literature can be exploited in the classroom and focus on the use of short stories as alternative materials that can encourage learners in a variety of classroom activities: from vocabulary enrichment to communicative abilities. Categories that encourage learners to develop the powers to interpret the texts are plot and suspense; characters and relationships; major themes; methods writer uses to communicate his attitudes; and reader’s response. Short stories can also lend themselves to intercultural values comparisons. Finally, it is expected that this paper can offer insights to other language teachers who are in similar settings.

  20. Dissection of genomic correlation matrices using multivariate factor analysis in dairy and dual-purpose cattle breeds

    Science.gov (United States)

    SNP effects estimated in genomic selection programs allow for the prediction of direct genomic values (DGV) both at genome-wide and chromosomal level. As a consequence, genome-wide (G_GW) or chromosomal (G_CHR) correlation matrices between genomic predictions for different traits can be calculated. ...

  1. 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.

  2. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-09-01

    This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.

  3. 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.

  4. 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.

  5. Packaging Design Criteria for the MCO Cask

    Energy Technology Data Exchange (ETDEWEB)

    FLANAGAN, B.D.

    2000-08-01

    Approximately 2,100 metric tons of unprocessed, irradiated, nuclear fuel elements are presently stored in the K Basins (including approximately 700 additional elements from the Plutonium-Uranium Extraction Plant, N Reactor, and 327 Laboratory). To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) 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 Multi-canister Overpacks. Concurrent with the K Basin cleanup, 72 Shippingport Pressurized Water Reactor Core 2 fuel assemblies will be transported from T Plant to the CSB to provide space at T Plant for K Basin sludge canisters.

  6. Packaging design criteria for the MCO cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1996-04-29

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins (including possibly 700 additional elements from PUREX, N Reactor, and 327 Laboratory). The basin water, particularly in the K East Basin, contains significant quantities of dissolved nuclear isotopes and radioactive fuel corrosion particles. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East area. In order to initiate K Basin cleanup on schedule, the two-year fuel-shipping campaign must begin by December 1997. 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.

  7. Packaging design criteria for the MCO cask

    Energy Technology Data Exchange (ETDEWEB)

    Clements, M.D.

    1996-01-01

    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.

  8. Packaging design criteria for the MCO cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1996-09-11

    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.

  9. 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.

  10. 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.

  11. Validation study for crediting chlorine in criticality analyses for spent nuclear fuel disposition

    Energy Technology Data Exchange (ETDEWEB)

    Sobes, Vladimir [ORNL; Scaglione, John M [ORNL; Wagner, John C [ORNL; Dunn, Michael E [ORNL

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of

  12. Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

    Energy Technology Data Exchange (ETDEWEB)

    Sobes, Vladimir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, Michael E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of

  13. Evolution of spent nuclear fuel during drying storage conditions; Evolucion del combustible nuclear gastado durante almacenamiento en condiciones de sequedad

    Energy Technology Data Exchange (ETDEWEB)

    Duro, L.; Riba, O.; Martinez-Esparza, A.; Bruno, J.

    2012-07-01

    The objective of this paper is a critical discussion of the main processes that determine the structure and radiological inventory CG UOX type and the very conditions of temporary storage. This study has allowed the configuration of state CG in potential function of oxygen affects the physico-chemical characteristics of irradiated fuel.

  14. REPROBING DNA BLOTS: WET IS BETTER THAN DRY STORAGE OF UNCHARGED NYLON MEMBRANES AFTER REMOVING PROBES. (R826602)

    Science.gov (United States)

    The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

  15. RE-PROBING DNA BLOTS: WET IS BETTER THAN DRY STORAGE OF UNCHARGED NYLON MEMBRANES AFTER REMOVING PROBES. (R826602)

    Science.gov (United States)

    The perspectives, information and conclusions conveyed in research project abstracts, progress reports, final reports, journal abstracts and journal publications convey the viewpoints of the principal investigator and may not represent the views and policies of ORD and EPA. Concl...

  16. Cell viability of microencapsulated Bifidobacterium animalis subsp. lactis under freeze-drying, storage and gastrointestinal tract simulation conditions.

    Science.gov (United States)

    Shamekhi, Fatemeh; Shuhaimi, Mustafa; Ariff, Arbakariya; Manap, Yazid A

    2013-03-01

    The purpose of this study was to improve the survival of Bifidobacterium animalis subsp. lactis 10140 during freeze-drying process by microencapsulation, using a special pediatric prebiotics mixture (galactooligosaccharides and fructooligosaccharides). Probiotic microorganisms were encapsulated with a coat combination of prebiotics-calcium-alginate prior to freeze-drying. Both encapsulated and free cells were then freeze-dried in their optimized combinations of skim milk and prebiotics. Response surface methodology (RSM) was used to produce a coating combination as well as drying medium with the highest cell viability during freeze-drying. The optimum encapsulation composition was found to be 2.1 % Na-alginate, 2.9 % prebiotic, and 21.7 % glycerol. Maximum survival predicted by the model was 81.2 %. No significant (p > 0.05) difference between the predicted and experimental values verified the adequacy of final reduced models. The protection ability of encapsulation was then examined over 120 days of storage at 4 and 25 °C and exposure to a sequential model of infantile GIT conditions including both gastric conditions (pH 3.0 and 4.0, 90 min, 37 °C) and intestinal conditions (pH 7.5, 5 h, 37 °C). Significantly improved cell viability showed that microencapsulation of B. lactis 10140 with the prebiotics was successful in producing a stable symbiotic powdery nutraceutical.

  17. Device for checking a plate accumulator battery suited for dry storage. Vorrichtung zur Pruefung eines trocken lagerfaehigen Plattenakkumulators

    Energy Technology Data Exchange (ETDEWEB)

    Krause, E.; Rust, G.; Ilgner, R.

    1986-10-23

    The invention concerns a device which permits the checking of dry plate batteries, where the plates are wetted with an electrolyte and therefore an off load measurement of the cell voltage is possible. This device consists of a paintbrush having normal bristles (without capillaries). In order to be able to carry small quantities of liquid (50 mg) on the bristles of the paintbrush, the paintbrush passes over a sponge, which is partly immersed in a reservoir of liquid. When placing the paintbrush on the plates to be measured, there is an ion-conducting bridge set up over the whole cell, so that the voltage produced by the cell can be measured.

  18. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [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); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Fermi Research Alliance (FRA), Batavia, IL (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 13 shutdown sites use designs from 4 different suppliers involving 11 different (horizontal and vertical) dry storage systems that would require the use of 9 different transportation cask designs to remove the SNF and GTCC waste from the shutdown sites. Although some changes to transportation certificates of compliance will be required, the SNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in dual purpose dry storage canisters that can be transported, including a small amount of high-burnup fuel. Most sites indicated that 2-3 years of advance time would be required for its preparations before shipments could begin. Some sites could be ready in less time. As additional sites such as Fort Calhoun, Clinton, Quad Cities, Pilgrim, Oyster Creek, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

  19. 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.

  20. 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.)

  1. Concepts for Small-Scale Testing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven Craig [Idaho National Laboratory; Winston, Philip Lon [Idaho National Laboratory

    2015-09-01

    This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

  2. DPC loading feasibility study report

    Energy Technology Data Exchange (ETDEWEB)

    Dafoe, R.E.; Lopez, D.A.; Williams, K.L.

    1997-11-01

    Disposal of radioactive wastes now stored at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory is mandated under a ``Settlement Agreement`` between the Department of Energy and the State of Idaho. This study investigates the feasibility of using the Dry Transfer Cell facility to package waste into Dual Purpose Canisters for interim storage at the adjacent Dry Storage System comprised of an interim storage pad with NUHOMS{reg_sign} storage modules. The wastes would then be road-ready for eventual disposal in a permanent repository. The operating period for these activities is expected to be from 2015 to 2035.

  3. 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.

  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. IKT-Gateway as a bridge between dual-purpose power plant and smart grid; IKT-Gateway als Bruecke zwischen BHKW und Smart Grid

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Joerg [SSV Software Systems GmbH, Hannover (Germany). Bereich Vertrieb und Marketing

    2012-07-15

    The way to a smart grid leads via the interconnection of distributed energy producers and consumers. Within the total network, energy producers and consumers have to communicate meaningfully with each other. This is a great performance for a reliable and performant long-distance service. In the case of 2G Energietechnik AG (Heek, Federal Republic of Germany), SSV Software Systems GmbH (Hannover, Federal Republic of Germany) has met this challenge with an ICT-based gateway solution mastered.

  6. Effect of a dual-purpose inoculant on the quality and nutrient losses from corn silage produced in farm-scale silos.

    Science.gov (United States)

    Queiroz, O C M; Adesogan, A T; Arriola, K G; Queiroz, M F S

    2012-06-01

    This project aimed to determine effects of applying an inoculant containing homofermentative and heterofermentative bacteria on the fermentation, nutritive value, aerobic stability, and nutrient losses from corn silage produced in farm-scale silos. Corn forage was harvested at 34% dry matter (DM) and treated without (control) or with 5 × 10⁵ cfu/g of Lactobacillus buchneri and Pediococcus pentosaceus. The inoculant was sprayed on alternate 8-row-wide swaths of forage, and the untreated and inoculated forages were alternately packed into 3.6-m-wide bag silos. Forty-five tonnes of corn forage were packed into each of 4 replicate bags per treatment and ensiled for 166 d. Silage removed from the bags (500 kg/d) was separated into good and spoiled (visibly moldy or darker) silage portions, and weighed for 35 d. Weekly composites were analyzed for chemical composition, aerobic stability, and fungal counts. Aerobic stability was measured using data loggers that recorded sample and ambient temperature every 30 min for 7 d. Inoculation did not affect the chemical composition of the spoiled or good silage but decreased the quantity (5.7 vs. 12.9 kg/d) and percentage (3.4 vs. 7.8) of spoiled silage in the bags by over 50%. Losses of crude protein (0.28 vs. 0.92 kg/d), gross energy (6.0 × 10⁴ vs. 1.8 × 10⁵ kJ/d), and neutral detergent fiber (1.34 vs. 4.12 kg/d) in spoiled silage were less in inoculated versus control silages. Inoculated silages had lower pH (3.91 vs. 3.99), lactate concentration (7.63 vs. 7.86%), lactate:acetate ratio (1.58 vs. 2.53%), and a greater acetate (5.11 vs. 3.56%) concentration than the control silage. Inoculated silages tended to have fewer yeasts (2.59 vs. 4.62 log cfu/g) than control silages, but aerobic stability was not different across treatments (14.7 vs. 9.5 h). Applying the inoculant made the fermentation more heterolactic, inhibited the growth of yeasts, and substantially reduced the amount of spoilage and the associated energy and nutrient losses.

  7. Shortening the postpartum anoestrous interval in suckled crossbred dual purpose cows using progestagen intravaginal sponges plus eCG and PGF(2alpha).

    Science.gov (United States)

    Gutiérrez, J C; Palomares, R; González, R; Portillo, G; Montero-Urdaneta, M; Rubio-Guillén, J; Hernández-Fonseca, H J; Soto-Belloso, E

    2009-02-01

    One hundred and twenty-six suckled crossbred cows (Bos taurus x Bos indicus), with body condition score >or=3 (1-5 point scale), were employed in the present study to evaluate the effectiveness of intravaginal progestin-releasing sponges (IVS) for shortening anoestrous interval. Fifty-four cows were assigned to control group. Seventy-two cows were treated with IVS impregnated with 250 mg of medroxy-acetate-progesterone (MAP) as follows: day 0, IVS plus 5 mg of 17beta-E and 50 mg of MAP i.m.; day 6, 500 IU of equine chorionic gonadotropin and 25 mg prostaglandin F(2alpha) i.m.; day 8, IVS withdrawal and day 9, 1 mg 17beta-E i.m. Cows were also grouped according to postpartum days (dpp) at treatment: MAP 70 days (n = 47); control >70 days (n = 32). From IVS removal, cows were detected in oestrus and inseminated. Cows not detected in oestrus were timed artificial insemination 72 h after sponge removal. Treatment effect on oestrous rate (ER), conception rate (CR), pregnancy rate (PR) and treatment to conception intervals (TCI) and calving to conception intervals (CCI) were evaluated. The ER, CR and PR were analysed using PROC LOGISTIC, while TCI and CCI with PROC GLM of SAS. The groups MAP 70 days showed higher (p 70 days (84.0% and 76.6% vs 31.8% and 31.3% respectively). The PR was higher (p 70 days vs control cattle.

  8. Design and experimental testing of air slab caps which convert commercial electron diodes into dual purpose, correction-free diodes for small field dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Charles, P. H., E-mail: paulcharles111@gmail.com [Department of Radiation Oncology, Princess Alexandra Hospital, Ipswich Road, Woolloongabba, Brisbane, Queensland 4102, Australia and School of Chemistry, Physics and Mechanical Engineering, Queensland University of Technology, GPO Box 2434, Brisbane, Queensland 4001 (Australia); Cranmer-Sargison, G. [Department of Medical Physics, Saskatchewan Cancer Agency, 20 Campus Drive, Saskatoon, Saskatchewan S7L 3P6, Canada and College of Medicine, University of Saskatchewan, 107 Wiggins Road, Saskatoon, Saskatchewan S7N 5E5 (Canada); Thwaites, D. I. [Institute of Medical Physics, School of Physics, University of Sydney, New South Wales 2006 (Australia); Kairn, T. [School of Chemistry, Physics and Mechanical Engineering, Queensland University of Technology, GPO Box 2434, Brisbane, Queensland 4001, Australia and Genesis CancerCare Queensland, The Wesley Medical Centre, Suite 1, 40 Chasely Street, Auchenflower, Brisbane, Queensland 4066 (Australia); Crowe, S. B.; Langton, C. M.; Trapp, J. V. [School of Chemistry, Physics and Mechanical Engineering, Queensland University of Technology, GPO Box 2434, Brisbane, Queensland 4001 (Australia); Pedrazzini, G. [Genesis CancerCare Queensland, The Wesley Medical Centre, Suite 1, 40 Chasely Street, Auchenflower, Brisbane, Queensland 4066 (Australia); Aland, T.; Kenny, J. [Epworth Radiation Oncology, 89 Bridge Road, Richmond, Melbourne, Victoria 3121 (Australia)

    2014-10-15

    Purpose: Two diodes which do not require correction factors for small field relative output measurements are designed and validated using experimental methodology. This was achieved by adding an air layer above the active volume of the diode detectors, which canceled out the increase in response of the diodes in small fields relative to standard field sizes. Methods: Due to the increased density of silicon and other components within a diode, additional electrons are created. In very small fields, a very small air gap acts as an effective filter of electrons with a high angle of incidence. The aim was to design a diode that balanced these perturbations to give a response similar to a water-only geometry. Three thicknesses of air were placed at the proximal end of a PTW 60017 electron diode (PTWe) using an adjustable “air cap”. A set of output ratios (OR{sub Det}{sup f{sub c}{sub l}{sub i}{sub n}}) for square field sizes of side length down to 5 mm was measured using each air thickness and compared to OR{sub Det}{sup f{sub c}{sub l}{sub i}{sub n}} measured using an IBA stereotactic field diode (SFD). k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} was transferred from the SFD to the PTWe diode and plotted as a function of air gap thickness for each field size. This enabled the optimal air gap thickness to be obtained by observing which thickness of air was required such that k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} was equal to 1.00 at all field sizes. A similar procedure was used to find the optimal air thickness required to make a modified Sun Nuclear EDGE detector (EDGEe) which is “correction-free” in small field relative dosimetry. In addition, the feasibility of experimentally transferring k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} values from the SFD to unknown diodes was tested by comparing the experimentally transferred k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} values for unmodified PTWe and EDGEe diodes to Monte Carlo simulated values. Results: 1.0 mm of air was required to make the PTWe diode correction-free. This modified diode (PTWe{sub air}) produced output factors equivalent to those in water at all field sizes (5–50 mm). The optimal air thickness required for the EDGEe diode was found to be 0.6 mm. The modified diode (EDGEe{sub air}) produced output factors equivalent to those in water, except at field sizes of 8 and 10 mm where it measured approximately 2% greater than the relative dose to water. The experimentally calculated k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} for both the PTWe and the EDGEe diodes (without air) matched Monte Carlo simulated results, thus proving that it is feasible to transfer k{sub Q{sub c{sub l{sub i{sub n,Q{sub m{sub s{sub r}{sup f{sub c}{sub l}{sub i}{sub n},f{sub m}{sub s}{sub r}}}}}}}}} from one commercially available detector to another using experimental methods and the recommended experimental setup. Conclusions: It is possible to create a diode which does not require corrections for small field output factor measurements. This has been performed and verified experimentally. The ability of a detector to be “correction-free” depends strongly on its design and composition. A nonwater-equivalent detector can only be “correction-free” if competing perturbations of the beam cancel out at all field sizes. This should not be confused with true water equivalency of a detector.

  9. Targhee Russet: A high yielding dual purpose, long russet potato cultivar having higher protein and vitamin C content and resistance to tuber soft rot

    Science.gov (United States)

    Targhee Russet is a dark-skinned russet potato variety with tubers slightly longer than Russet Burbank. It produces higher total and marketable yields than does Russet Burbank at most of the sites it was tested in the western United States. Tuber dormancy is about 58 days shorter than Russet Burba...

  10. Preliminary assessment of radiological doses in alternative waste management systems without an MRS facility

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Pelto, P.J.; Daling, P.M.; Lavender, J.C.; Fecht, B.A.

    1986-06-01

    This report presents generic analyses of radiological dose impacts of nine hypothetical changes in the operation of a waste management system without a monitored retrievable storage (MRS) facility. The waste management activities examined in this study include those for handling commercial spent fuel at nuclear power reactors and at the surface facilities of a deep geologic repository, and the transportation of spent fuel by rail and truck between the reactors and the repository. In the reference study system, the radiological doses to the public and to the occupational workers are low, about 170 person-rem/1000 metric ton of uranium (MTU) handled with 70% of the fuel transported by rail and 30% by truck. The radiological doses to the public are almost entirely from transportation, whereas the doses to the occupational workers are highest at the reactors and the repository. Operating alternatives examined included using larger transportation casks, marshaling rail cars into multicar dedicated trains, consolidating spent fuel at the reactors, and wet or dry transfer options of spent fuel from dry storage casks. The largest contribution to radiological doses per unit of spent fuel for both the public and occupational workers would result from use of truck transportation casks, which are smaller than rail casks. Thus, reducing the number of shipments by increasing cask sizes and capacities (which also would reduce the number of casks to be handled at the terminals) would reduce the radiological doses in all cases. Consolidating spent fuel at the reactors would reduce the radiological doses to the public but would increase the doses to the occupational workers at the reactors.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. Obstruction of Water Uptake in cut Chrysanthemum Stems after Dry Storage: Role of Wound-induced Increase in Enzyme Activities and Air Emboli

    NARCIS (Netherlands)

    Meeteren, van U.; Arevalo-Galarza, L.

    2009-01-01

    Hydraulic conductance of cut chrysanthemum stems was lowered by the aspiration of air as well as by a wound-induced plant response. By measuring the hydraulic conductance of stem segments in which air could be introduced into and/or removed from the xylem vessels at various times after harvest, we

  16. Evaluating manta ray mucus as an alternative DNA source for population genetics study: underwater-sampling, dry-storage and PCR success

    Directory of Open Access Journals (Sweden)

    Tom Kashiwagi

    2015-08-01

    Full Text Available Sharks and rays are increasingly being identified as high-risk species for extinction, prompting urgent assessments of their local or regional populations. Advanced genetic analyses can contribute relevant information on effective population size and connectivity among populations although acquiring sufficient regional sample sizes can be challenging. DNA is typically amplified from tissue samples which are collected by hand spears with modified biopsy punch tips. This technique is not always popular due mainly to a perception that invasive sampling might harm the rays, change their behaviour, or have a negative impact on tourism. To explore alternative methods, we evaluated the yields and PCR success of DNA template prepared from the manta ray mucus collected underwater and captured and stored on a Whatman FTA™ Elute card. The pilot study demonstrated that mucus can be effectively collected underwater using toothbrush. DNA stored on cards was found to be reliable for PCR-based population genetics studies. We successfully amplified mtDNA ND5, nuclear DNA RAG1, and microsatellite loci for all samples and confirmed sequences and genotypes being those of target species. As the yields of DNA with the tested method were low, further improvements are desirable for assays that may require larger amounts of DNA, such as population genomic studies using emerging next-gen sequencing.

  17. Simplified model of the thermal evolution of the fuel dry storage; Modelo simplificado de la evolucion termica del combustible en almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Penalva, J.; Feria, F.; Herranz, L. E.

    2013-07-01

    The objective of this work is to establish a model that allows to determine easily temperature of the fuel burned and time out of reactor. The model will depend on the selected storage system, but the established methodology lays the groundwork for your application to any other case. to obtain the temperature has been used the FLUENT code, based on a model 3D itself and in the heat of decay at different burn (30-60 MWd/kgU) calculated with origin (information available in the literature). Of the different simulations carried out has developed a 3D map of temperature versus burned and time out of reactor, which has allowed to develop an equation simplified.

  18. Polysaccharide benefits dry storage survival of the biocontrol agent Pseudomonas fluorescens S11:P:12 effective against several maladies of stored potatoes

    Science.gov (United States)

    Pseudomonas fluorescens S11:P:12 (NRRL B-21133) is a biological control agent able to suppress several storage maladies of potatoes including sprouting, Fusarium dry rot incited by Gibberella pulicaris, pink rot incited by Phytophthora erythroseptica, and late blight incited by Phytophthora infestan...

  19. Polysaccharide Production Benefits Dry Storage Survival of the Biocontrol Agent Pseudomonas fluorescens S11:P:12 Effective Against Several Maladies of Stored Potatoes

    Science.gov (United States)

    Pseudomonas fluorescens S11:P:12 (NRRL B-21133) is a biological control agent able to suppress several potato diseases and sprouting. Notably, it produces a polysaccharide during liquid cultivation; and the objective of this work was to determine the role of this material in the bio-control process...

  20. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  1. 78 FR 73379 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ... Federal and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory Commission... February 19, 2014, unless significant adverse comments are received by January 6, 2014. If the direct final... and Management System (ADAMS): You may access publicly available documents online in the NRC Library...

  2. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ... Federal and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory Commission... editorial corrections. DATES: Submit comments by January 6, 2014. Comments received after this date will be...-0052. NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly...

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

  8. Assembling a dual purpose TaqMan-based panel of single-nucleotide polymorphism markers in rainbow trout and steelhead (Oncorhynchus mykiss) for association mapping and population genetics analysis

    DEFF Research Database (Denmark)

    Hansen, Mette H H; Young, Sewall; Jørgensen, Hanne Birgitte Hede;

    2011-01-01

    We establish a TaqMan-based assay panel for genotyping single-nucleotide polymorphisms in rainbow trout and steelhead (Oncorhynchus mykiss). We develop 22 novel single-nucleotide polymorphism markers based on new steelhead sequence data and on assays from sister taxa. Additionally, we adapt 154...... previously developed markers to the TaqMan platform. At the beginning of this study, 59 SNPs with TaqMan assays were available to the scientific community. By adding 176 additional TaqMan assays to this number, we greatly expand the biological applications of TaqMan genotyping within both population genetics...... and quantitative genetics...

  9. Avaliação de esquemas de seleção alternativos para bovinos Zebus de dupla aptidão Evaluation of alternative breeding programs for dual purpose Zebu cattle

    Directory of Open Access Journals (Sweden)

    Raimundo Nonato Braga Lôbo

    2000-10-01

    Full Text Available Utilizando-se simulação determinística, foram avaliados seis esquemas de melhoramento, com diferentes grupos de informações: (1 somente características de crescimento, (2 somente características leiteiras, (3 características de crescimento e reprodutivas, (4 características leiteiras e reprodutivas, (5 características de crescimento e leiteiras e (6 características de crescimento, leiteiras e reprodutivas. Avaliaram-se também a resposta à variação no uso de touros provados por meio do teste de progênie, a resposta sem a realização do teste de progênie no estrato de vacas de corte da população e a resposta à variação no valor econômico do peso ao abate. Em todos os esquemas avaliados, foram verificados altos ganhos genéticos e lucros decorrentes do programa. Os esquemas que consideraram as características leiteiras foram mais eficientes pela seleção para maior produção de leite e menor peso de vaca adulta. As características de fertilidade foram mais importantes para os esquemas de corte. A resposta econômica à seleção foi maior com o aumento no uso de touros provados, embora não proporcionalmente. A contribuição das características de corte para o lucro genético somente superou a das leiteiras quando o valor econômico do peso ao abate superou em 14 vezes o da produção de leite. Foi demonstrado não haver necessidade da realização do teste de progênie com as vacas de corte para o sistema de dupla aptidão estudado.A deterministic simulation was used to compare six breeding schemes based on different levels of performance recording: (1 only body weights, (2 only dairy traits, (3 body weights and reproductive traits, (4 dairy and reproductive traits, (5 body weights and dairy traits and (6 body weights, reproductive and dairy traits. Other factors were also studied, as the level of usage of proven sires, variations in the economic value of slaughter weight and eliminating progeny testing in the beef herd tier of the population. Genetic gains and returns were high in all schemes. Dairy schemes were superior, mainly due to responses in lowering mature cow weight and increasing milk yield. Reproductive traits were more important in the beef schemes. Returns increased with higher usage of proven sires, although not proportionally. The contribution of beef traits to genetic profit was higher than that of dairy traits only when the economic value of slaughter weight exceeded 14 times the economic value of milk yield. The exclusion of progeny testing in the beef cows tier had little effect on the genetic or economic results.

  10. Producción de leche de vacas mestizas del Sistema Doble Propósito en el municipio de Arauca -Milk production of cows crossbred dual purpose system of municipalities of Arauca.

    Directory of Open Access Journals (Sweden)

    Salamanca, CA

    2012-07-01

    Full Text Available ResumenLa producción de leche y la duración de la lactancia son parámetros que permite conocer la cantidad de leche producida por la vaca y por finca.AbstractMilk production and lactation length are parameters to find out the amount of milk produced by cows and per farm.

  11. Projected dual-purpose use of a rectangular bolted section at a depth of 1200 m in the D2/C seam; Geplante Doppelnutzung einer Rechteckankerstrecke in 1200m Teufe im Floez D2/C

    Energy Technology Data Exchange (ETDEWEB)

    Breinig, F.; Opolony, K. [Deutsche Steinkohle AG (Germany)

    2004-07-01

    The report gives an insight view to the planning of excavation and support design in seam D2/C at the Lippe mine. Starting with the first rectangular rockbolted tailgate from panel 328 and the subproject rectangular rockbolted roadways for advanced mining'' up to the design of rockbolting pattern for re-use of the maingate of panel 330 in the D2/C seam. The engineering and development of a combined support and casing-unit for recent roadside packages are descibed as well as experience in new dimensioning based upon numerical and physical modelling. Finally the operational experience of the first use are displayed and analysed.

  12. Structure and Application Methods of Large Dual-purpose (Light & Heavy) Compaction Device%轻重两用型大型击实仪的仪器结构和使用方法

    Institute of Scientific and Technical Information of China (English)

    尤苏南

    2012-01-01

    普通标准击实仪不适用于含大粒径粗粒土,野外施工有时没有动力电源。为了满足工程需要,研制了一种轻重两用型大型击实仪,仪器主要由底板、击实筒、穿心击实锤、穿心抓体、落锤高度控制装置和其它辅助装置组成。击实仪的击实筒装置旋转,提锤和落锤装置不旋转,提锤并自动控制落锤部件采用自动脱钩自由落锤,提锤动力采用柴油发动机勘察钻机卷扬。实际使用结果表明,能满足大型击实试验的需要,有一定的工程应用价值。%For common standared compaction device is not suitable for large-size coarse-grained soil,and once there is no power during field construction,in order to satisfy the needs of construction,a large dual-purpose(light heavy) compaction device is developed and produced,which is composed of baseboard,compaction cylinder,cross-core compaction hammer,cross-core catch body,control device of drop-hammer height and other accessory devices.The compaction cylinder is able to rotate,but the lift hammer and drop hammer are not.Automatic decoupling and free drop hammer are used for the parts lifting the hammer and controlling the drop hammer freely.The lift hammer could be winded by surveying rig with diesel motor.The actual application results show that this compaction device can satisfy the needs of large compaction tests and has a certain engineering application value.

  13. Evaluación genética del recurso animal de los sistemas de producción de bovinos en doble propósito en Colombia / Genetic evaluation of dual purpose production systems in Colombia

    OpenAIRE

    Galeano Rivera, Adriana Patricia

    2010-01-01

    Con el objetivo de determinar el verdadero potencial genético del recurso animal disponible en los sistemas de producción bovina de doble propósito del trópico bajo colombiano, se evaluaron los registros de producción de leche por lactancia (Kg), peso al destete (Kg), intervalo entre partos (d) e Índice de Vaca, de 1.687 hembras reproductoras, durante los años 1998 y 2007. Se empleó un modelo animal mixto que incluyó los efectos genéticos aleatorios del animal, el medio ambiente permanente y ...

  14. 不同消毒方法在管道分质供水工程中的应用比较%Comparison of different disinfection techniques for dual purpose pipe water supply system

    Institute of Scientific and Technical Information of China (English)

    吴贤格; 肖贤明; 杨毅; 陈启华; 罗冬浦

    2008-01-01

    管道分质供水常用的消毒方法有臭氧法、紫外(UV)法、TCAA法、二氧化氯(ClO2)法及组合工艺法等,对不同消毒方法在管道分质供水工程中的应用情况进行比较研究.结果表明,在不同消毒方法中,应用TCCA+UV与CIO2+UV组合工艺法,只要控制用户终端净水中的消毒剂浓度在0.05 mg/L时,对管道分质供水消毒具有较好的效果,同时可以保持水质具有良好的口感.

  15. Fasciola hepatica en bovinos doble propósito de una finca de Tucacas (Falcón, Venezuela - Fasciola hepatica in a dual-purpose bovine herd from a cattle farm in Tucaras (Falcón, Venezuela

    Directory of Open Access Journals (Sweden)

    Pérez-Mata, Arlett

    2009-04-01

    Full Text Available ResumenLa fasciolosis es una enfermedad parasitaria causada por eltrematodo Fasciola hepatica, la cual causa pérdidas económicas pordecomiso de hígados en mataderos y disminución en la producción deleche y carne. Esta investigación tuvo dos objetivos: 1. Determinar laprevalencia de Fasciola hepatica en ganado bovino doble propósitoproveniente de una finca de Tucacas, Municipio Silva, Estado Falcón,Venezuela, 2. Determinar la presencia de los hospedadoresintermediarios del parásito en dicha finca. En el 2006, se realizarondos muestreos en animales adultos en edad productiva: El primero alinicio de la época seca (Época 1 = E1, en el mes de enero,utilizándose el 10 % (n = 30 de la población. El segundo se efectuóen la mitad de esa época (Época 2 = E2, en el mes de abril con el10% (n = 18 de la población. En ambas ocasiones se tomaronmuestras coprológicas individuales que fueron procesadas por laTécnica de Sedimentación. Se tomaron muestras sanguíneas paradeterminar el hematocrito. Los resultados indican una prevalencia deFasciola hepatica del 20% en la E1 y del 11,11% en la E2(p>0,05.Se encontraron caracoles lymnaeidos en 4 potreros inspeccionados enla E1, y sólo en uno de ellos en la E2. Se comprobó la infección porFasciola hepatica sólo en los caracoles de un potrero en la E1. Seconcluye que el parasitismo es endémico en la finca no habiendodiferencias estadísticamente significativas en la prevalencia deltrematodo en sus hospedadores bovinos entre las dos épocas. Lapoblación de caracoles en los potreros disminuyó significativamente(p0,05. Lymnaeid snails were found in alllots searched (n = 4 during S1 and only in one lot during S2.Infection with Fasciola hepatica was detected only in snails from one of the lots, during S1. We conclude that this parasitic disease became endemic in the farm. There were not statistical differences inprevalence in definitive bovine hosts between S1 and S2, but snailpopulation decreased (p < 0,05 between these two seasons from214 (S1 to 18 snails (S2.

  16. 青海毛肉兼用半细毛羊生产性能遗传力估测%Estimation of Heritability of Performance for Qinghai Semifinal Wool sheep with Dual-purpose Type of Wool and Meat

    Institute of Scientific and Technical Information of China (English)

    李楼太; 高玉兰; 刘玉国

    2009-01-01

    通过对青海省柴达木地区青海半细毛羊9个主要生产性能进行,调查研究分析.结果表明:初生重、断奶重、周岁剪毛量、羊毛细度、毛长、剪毛前体重、剪毛量都是高遗传力的生产性能,也是青海半细毛羊育种过程中鉴定选种、组群的主要选择依据;断奶重、周岁剪毛量、剪毛前体重与剪毛量之间呈强正遗传相关,断奶重可作为对剪毛量进行早期选择的生产性能指标;断奶重与羊毛细度之间呈负遗传相关.

  17. 基于ADAMS的某型通用机枪动力学建模与仿真%Virtual Prototyping of Dual-purpose Machine Gun Based on ADAMS and Simulation

    Institute of Scientific and Technical Information of China (English)

    陈明; 马吉胜; 贾长治; 张建宇; 林奎

    2006-01-01

    在分析某型通用机枪射击时运动和受力情况的基础上,利用PRO/E和ADAMS软件建立了该机枪的虚拟样机模型,并对其进行了校核,结果证明该虚拟样机与机枪的运动和受力情况基本一致,符合工程分析的要求.在此基础上,对机枪射击时的动态特性进行了仿真计算,研究结果说明基于虚拟样机技术进行机枪动力学特性分析是可行的.

  18. 基于虚拟样机的某型机枪射击动态性能研究%Firing Dynamic Characteristics of a Dual-purpose Machine Gun Based on Virtual Prototype

    Institute of Scientific and Technical Information of China (English)

    陈明; 马吉胜; 贾长治; 张建宇

    2006-01-01

    利用Pro/E和ADAMS建立了某型通用机枪射击时的多体系统动力学模型.将内弹道计算所得的发射时的外载荷施加于相应位置,针对轻机枪和重机枪两种不同状态对其射击时的动力学特性进行仿真分析,得到机枪在五连发射击时关健部位的载荷和运动规律.以枪口跳动为指标,研究了复进簧刚度、俯仰刚度和阻尼、回转刚度和阻尼、枪机质量偏心、枪机框质量偏心等结构参数对机枪射击精度的影响规律.该文结果可为机枪的结构优化提供参考.

  19. EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

  20. 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)

  1. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David; Bryan, Charles R.

    2015-10-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  2. ¬¬¬¬ SURVIVAL OF Cronobacter sakazakii IN SKIM MILK DURING SPRAY DRYING, STORAGE AND RECONSTITUTION [Ketahanan Hidup Cronobacter sakazakii dalam Susu Skim selama Proses Pengeringan Semprot, Penyimpanan dan Rekonstitusi

    Directory of Open Access Journals (Sweden)

    Lilis Nuraida1,2

    2012-12-01

    Full Text Available Cronobacter sakazakii is an emerging pathogen known to survive dry conditions and its presence in powder infant formula (PIF has been linked to several outbreaks. In Indonesia, isolation of this bacterium from various foods have been reported. The objective of this study was to determine the effect of spray drying and storage humidity on the survival of C. sakazakii YRc3a in skim milk and their viability upon reconstitution. The survival of Cronobacter during spray drying was determined by comparing the number of bacteria before and after drying. The viability of Cronobacter in spray dried skim milk (SDSM during storage was observed at weeks 1 to 8 and 12. At the same intervals, SDSM containing the pathogens was reconstituted at either 27°C or 50°C and the survivors were enumerated. The data were plotted to yield survival curves. Spray drying caused 4.19 log CFU/g reduction of Cronobacter and the bacteria experiencing drying were less sensitive to reconstitution at 50°C. During storage, the water activity of SDSM reached equilibrium at week 2 and afterwards, they started to decrease when stored at 50% or 90% RH, but maintained its viability at 70% RH. Storage at 50% and 90% RH accelerated the death rate of C. sakazakii YRc3a, resulting in the decline of the viable counts for 3 log cycles. At 50% RH, C. sakazakii Yrc3a decreased significantly, but the survivors exhibited increased heat resistance with the lowest reduction upon reconstitution at 50°C (0.16 log CFU/ml.

  3. Study of the oxidation of the matrix of irradiated fuel under conditions of dry storage; Estudio de la oxidacion de la matriz del combustible irradiado en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Elorrieta, J. M.; Bonales, L. J.; Rodriguez, N.; Gutierrez, L.; Cobos, J.; Baonza, V. G.

    2014-07-01

    Spain holds the open cycle for the management of spent fuel of nuclear power stations. The solution adopted, given the urgent need for greater storage capacity, has been the construction of a centralized temporary storage, with an expected 100-year life, which will be stored from 2017 both spent fuel from nuclear power plants and other waste high activity of NPPs. (Author)

  4. 重水堆核电厂乏燃料干式中间贮存现状和技术%Status and Technology of Interim Spent Fuel Dry Storage Facility for PHWR Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    郑利民; 申森

    2005-01-01

    乏燃料干式贮存经过近30年的研发和改进已成为一种成熟的技术,乏燃料干式贮存总量正在显著增加.本文概要介绍重水堆核电厂乏燃料干式中间贮存的现状和技术,同时,提出秦山三期重水堆核电厂采用乏燃料干式中间贮存技术的初步设想.

  5. ¬¬¬¬ SURVIVAL OF Cronobacter sakazakii IN SKIM MILK DURING SPRAY DRYING, STORAGE AND RECONSTITUTION [Ketahanan Hidup Cronobacter sakazakii dalam Susu Skim selama Proses Pengeringan Semprot, Penyimpanan dan Rekonstitusi

    OpenAIRE

    2012-01-01

    Cronobacter sakazakii is an emerging pathogen known to survive dry conditions and its presence in powder infant formula (PIF) has been linked to several outbreaks. In Indonesia, isolation of this bacterium from various foods have been reported. The objective of this study was to determine the effect of spray drying and storage humidity on the survival of C. sakazakii YRc3a in skim milk and their viability upon reconstitution. The survival of Cronobacter during spray drying was determined by c...

  6. 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.

  7. 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...

  8. 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:...

  9. 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; ]...

  10. 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....

  11. 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...

  12. 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.

  13. 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.

  14. 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...

  15. 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...

  16. 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.

  17. 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.

  18. 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

  19. 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...

  20. 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...

  1. 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

  2. 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....

  3. Nuclear Industry Input to the Development of Concepts for the Consolidated Storage of Used Nuclear Fuel - 13411

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Chris; Thomas, Ivan; McNiven, Steven [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States); Lanthrum, Gary [NAC International, 3930 East Jones Bridge Road, Norcross, GA, 30092 (United States)

    2013-07-01

    EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the

  4. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  5. In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); De Baere, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Vaccaro, S. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Schwalbach, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management Company (Sweden); Tobin, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-01

    Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

  6. 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

    This Integrated Research Project (IRP) was established to characterize key limiting phenomena related to the performance of used nuclear fuel (UNF) storage systems. This was an applied engineering project with a specific application in view (i.e., UNF dry storage). The completed tasks made use of a mixture of basic science and engineering methods. The overall objective was to create, or enable the creation of, predictive tools in the form of observation methods, phenomenological models, and databases that will enable the design, installation, and licensing of dry UNF storage systems that will be capable of containing UNF for extended period of time. The project was divided four distinct, yet synergistic, technical mission areas (TMAs), as summarized below. The key technical results and findings from each of the TMAs are summarized in Sections 2 through 5. Technical Mission Area 1: Low Temperature Creep This mission focused on the low temperature creep of UNF cladding that may be enabled by decay heat from fission products and stresses from internal pressures. The major objectives were (1) to obtain data using highly oxidized/hydrided tubing under relevant stresses and temperatures, (2) to characterize and translate that data to enable input to FRAPCON and other codes that may be modified to predict UNF behavior in dry storage, and (3) to formulate atomistic simulations to better understand long term creep behavior. Technical Mission Area 2: Hydrogen Behavior and Delayed Hydride Cracking This mission focused on the characterization and understanding of delayed hydride cracking (DHC) in spent Zircaloy cladding. The DHC mechanism is generally attributed to local hydride precipitation at stress risers present on the surface of the cladding. Samples with low and high hydrogen loadings were prepared and studiedusing various methods. Technical Mission Area 3: UNF Canister Corrosion This mission was focused on recognized gaps in understanding mechanisms relevant to the

  7. 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

  8. The effect of single and double quenching heat treatments on the mechanical properties of low alloy steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeongpil; Kim, Byung Jun; Lee, Sanghui; Jang, Jaeho; Nam, Daegeun [Korea Institute of Industrial Technology, Busan (Korea, Republic of); Sung, Gichan [Sehwa M and P Co., LTD, Ulsan (Korea, Republic of)

    2014-05-15

    Dry cask storage is a method of storing high-level radioactive waste, such as spent nuclear fuel that has already been cooled in the spent fuel pool for at least one year. For spent fuel storage, low alloy steel is widely used for shielding materials for dry storage cask of spent fuel because of their excellent mechanical properties, weldability and low price. However, they may suffer embrittlement by high levels of radiation and heat for a long period. Therefore, it is important to improve mechanical properties of low alloy steel for the integrity of structure materials. Generally, conventional single quenching and tempering (CQT) heat treatment process is used to improve the mechanical properties by controlling the temperature and time of heat treatment. In this study, the microstructure and mechanical properties of DQT heat treated specimens were investigated to improve the mechanical properties of low alloy steels comparing the CQT heat treated specimen. The following conclusions were obtained. The grain size of DQT process has a fine average grain size comparing the CQT process. A fine grain structure with DQT specimen affects the mechanical properties such as the reduction of hardness and increments of elongation. The DBTT after DQT process is shifted to lower temperatures. Especially, the DBTT of steel-3 is shifted to lower temperature about -40 .deg. C comparing the CQT specimen(Steel-1). The reason for the reduction of DBTT and USE after the DQT process is related to the changes of microstructure which are transformed from bainite to pearlite phase and the reduction of grain size with decreasing the temperature of heat treatment.

  9. Ensilage performance of sorghum hybrids varying in extractable sugars

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, Dirk; Moore, Kenneth J. [Iowa State University, Ames (United States); Pedersen, Jeffrey F.; Mitchell, Robert B. [University of Nebraska, Lincoln, NE (United States). USDA-ARS, Department of Agronomy; Grant, Richard J. [William H. Miner Agricultural Research Institute, Chazy, NY (United States); Redfearn, Daren D. [Oklahoma State University, Stillwater, OK (United States). Department of Plant and Soil Sciences

    2007-07-15

    Renewable feedstock resources require novel storage technologies to optimize industrial use. Solid state fermentation of biomass feedstock may provide organic chemicals and fibers while reducing the risk of current dry-storage procedures. Here, we compare the chemical composition and fermentation of six sorghum hybrids (Sorghum bicolor L. Moench) following 1, 7, and 21 days of storage. Ensilage of 7 days resulted in a pH of 3.8 and declined further to 3.75 at day 21. Lactate increased during ensilage from 2.0 to 3.9 g 100 g{sup -1}. Acetic acid increased between 1 and 7 days of ensiling but did not change until the end of the ensiling period. Total organic acids averaged 2.5 g 100 g{sup -1} after day 1 and increased to 4.2 and 4.7 g '100 g{sup -1} after days 7 and 21, respectively. Neutral detergent fiber ranged from 38 to 50 g 100 g{sup -1} among hybrids and total non-structural carbohydrates varied from 18 to 32 g 100 g{sup -1}. Hemicellulose and cellulose ranged from 13 to 19 g 100 g{sup -1} and 20 and 28 g 100 g{sup -1}, respectively. Genotypic variation in sorghum may offer designing dual-purpose hybrids for production of biomass and economically valuable byproducts. (author)

  10. 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.

  11. Used Fuel Logistics: Decades of Experience with transportation and Interim storage solutions

    Energy Technology Data Exchange (ETDEWEB)

    Orban, G.; Shelton, C.

    2015-07-01

    Used fuel inventories are growing worldwide. While some countries have opted for a closed cycle with recycling, numerous countries must expand their interim storage solutions as implementation of permanent repositories is taking more time than foreseen. In both cases transportation capabilities will have to be developed. AREVA TN has an unparalleled expertise with transportation of used fuel. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation model that was initially developed in the 1970s has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous “lessons learned” have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. AREVA TN has also an extensive experience with interim dry storage solutions in many countries on-site but also is working with partners to developed consolidated interim storage facility. Both expertise with storage and transportation contribute to safe, secure and smooth continuity of the operations. This paper will describe decades of experience with a very successful transportation program as well as interim storage solutions. (Author)

  12. 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 ...

  13. 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.

  14. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  15. 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.

  16. 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)

  17. 德系西门塔尔牛与荷斯坦牛产奶性状及血清细胞因子比较研究%Comparative Research of Milk Production Traits and Serum Cytokines Between Dual Purposes of Simmental and Holstein

    Institute of Scientific and Technical Information of China (English)

    张志超; 王雅春; 俞英

    2016-01-01

    本研究对来自河北省张北地区同一牛场的德系西门塔尔牛和荷斯坦牛的产奶性状、体细胞数和血清因子水平进行了检测和对比分析.结果表明:德系西门塔尔牛平均日产奶量15.7 kg、年平均产奶量4.7 t,同场荷斯坦牛分别为21.6 kg/d、6.5 t/年,但德系西门塔尔牛平均乳蛋白率(3.57%)极显著高于荷斯坦牛(3.16%)(P<0.01),乳脂率也明显高于荷斯坦牛(P=0.09).此外,德系西门塔尔平均体细胞数为8.5万/mL,明显低于全国平均水平(46.2万/mL),表现出良好的乳房健康状况.德系西门塔尔牛血清中TRAPPC9及IL-17水平均显著高于荷斯坦牛(P<0.05).相关分析显示,德系西门塔尔牛乳蛋白率与TRAPPC9含量显著相关(P<0.05),IL-17与TNF-α、IFN-γ显著相关(P<0.05).以上数据表明,引进德系西门塔尔牛在我国张北等寒冷地区具有较好适应性,乳房健康状况良好,产奶性能优良.

  18. Estimación de polimorfismos del gen de leptina de sementales en el sistema doble proposito bovino, en villaflores, Chiapas, México - Estimation of leptin gene polymorphisms in sire into the dual purpose bovine system, in villaflores, Chiapas, México

    Directory of Open Access Journals (Sweden)

    Ruíz Sesma, Benigno

    2009-12-01

    Full Text Available ResumenLa estimación del polimorfismos del gen leptina incorporado a losprogramas de selección asistidas por marcadores puede hacer máseficientes los sistemas de producción animal. La mutación del genleptina TT esta asociado con la eficiencia alimenticia y calidad de lacarne en bovinos. El estudio se realizó en el municipio de Villaflores,Chiapas, México. El objetivo del estudio fue estimar las frecuenciasgenotípicas y alélicas del gen leptina en el exon 2. El polimorfismo fueobtenido mediante un análisis molecular con la técnica ARMS-PCR. Seestudiaron 47 sementales en servicio activo en los sistemas deproducción bovina de doble propósito. Los resultados mostraron que el74.5 % de los sementales fueron Suizo Americano, 6.4 % Holstein y el12.8 % cruzas raciales. Con respecto al polimorfismo del gen leptina el19 % de sementales presentaron el genotipo TT, 32 % CC y 49 % TC.Los sementales de las diferentes razas o cruzas presentaron mayorfrecuencia de CT que de TT y CC ( χ2, P>0.05. Se concluye que la bajafrecuencia genotípica TT es debido probablemente a que los productoresen este sistema seleccionan sus reproductores basados en su tipo racial,sin considerar caracteres de producción de leche o calidad de la carneque se asocian con la mutación del gen leptina.SummaryLeptin gene polymorphisms incorporated in markers assisted selectionprograms could to make more efficient the animal production systems.The leptin gene mutation TT is associating to feed efficient and meatquality in cattle. The study was carried out in the municipality ofVillaflores, Chiapas. The objective was determinate genotypic and allelicfrequencies of the leptin gene in the exon 2. The polymorphism wasobserved by ARMS-PCR. 47 sires in active service into the dual purposebovine production system were included in this study. 74.5 % of sireswere Brown Swiss, 6.4% Holstein, and 12.8 % crosses of differentbreeds. The results indicated for sires, 19 % genotype TT, 32 %genotype CC, and 49 % genotype TC. In the group of sires crosses ofdifferent breeds had higher CT genotype frequency ( χ2, P>0.05. It wasconcluded that a low frequency of TT genotype could be associated abad selection criterion by producers due they select their breedinganimals based on racial type. Then, they not are considering milkproduction or meat quality traits.

  19. 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...

  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. 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.

  2. 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.

  3. 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.

  4. 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...

  5. Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Leotlela, Mosebetsi J. [Witwatersrand Univ., Johannesburg (South Africa). School of Physics; Koeberg Operating Unit, Johannesburg (South Africa). Regulations and Licensing; Malgas, Isaac [Koeberg Nuclear Power Station, Duinefontein (South Africa). Nuclear Engineering Analysis; Taviv, Eugene [ASARA consultants (PTY) LTD, Johannesburg (South Africa)

    2015-11-15

    In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k{sub eff}. Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

  6. 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

  7. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    L.L. Swanson

    2005-04-06

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts

  8. 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

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  9. 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

  10. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

  11. 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

  12. 《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.

  13. 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...

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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 abse