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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to...

78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI...

Status update of the BWR cask simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel G.

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 maximummore » 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 of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on heat load and the effect of simulated wind on a simplified below ground vent configuration.« less

75 FR 27463 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1; Correction AGENCY: Nuclear Regulatory... fuel storage casks to add revision 1 to the NUHOMS HD spent fuel storage cask system. This action is... Federal Register on May 7, 2010 (75 FR 25120), that proposes to amend the regulations that govern storage...

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

NASA Astrophysics Data System (ADS)

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 developed and calculated. Main criteria for estimating the maximum leakage rate for the lid metallic seal system are no loss of the pre-stress of the lid bolts, no appearance of the plastic region between the metal seal flanges, and no large relative deformation of the lid seals. Finally, in both cases, the low leakage rate for the metal cask lid closure system under the impulsive loads due to aircraft engine crash will be proved thoroughly.

Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael Tyacke; Frantisek Svitak; Jiri Rychecky

2010-04-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design,more » licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.« less

Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

2007-10-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing,more » testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Lindgren, Eric Richard

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 also by increasing themore » internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.« less

Test Plan for the Boiling Water Reactor Dry Cask Simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel; Lindgren, Eric R.

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 themore » 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 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 of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.« less

78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-16

... Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections AGENCY: Nuclear Regulatory Commission... revising the Holtec International, Inc. (Holtec) HI-STORM 100 Cask System listing within the ``List of... the Holtec HI-STORM 100 Cask System, Amendment No. 8. The purpose of this document is to provide...

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

Evaluation of RAPID for a UNF cask benchmark problem

NASA Astrophysics Data System (ADS)

Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

2017-09-01

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

Adapting Dry Cask Storage for Aging at a Geologic Repository

DOE Office of Scientific and Technical Information (OSTI.GOV)

C. Sanders; D. Kimball

2005-08-02

A Spent Nuclear Fuel (SNF) Aging System is a crucial part of operations at the proposed Yucca Mountain repository in the United States. Incoming commercial SNF that does not meet thermal limits for emplacement will be aged on outdoor pads. U.S. Department of Energy SNF will also be managed using the Aging System. Proposed site-specific designs for the Aging System are closely based upon designs for existing dry cask storage (DCS) systems. This paper evaluates the applicability of existing DCS systems for use in the SNF Aging System at Yucca Mountain. The most important difference between existing DCS facilities andmore » the Yucca Mountain facility is the required capacity. Existing DCS facilities typically have less than 50 casks. The current design for the aging pad at Yucca Mountain calls for a capacity of over 2,000 casks (20,000 MTHM) [1]. This unprecedented number of casks poses some unique problems. The response of DCS systems to off-normal and accident conditions needs to be re-evaluated for multiple storage casks. Dose calculations become more complicated, since doses from multiple or very long arrays of casks can dramatically increase the total boundary dose. For occupational doses, the geometry of the cask arrays and the order of loading casks must be carefully considered in order to meet ALARA goals during cask retrieval. Due to the large area of the aging pad, skyshine must also be included when calculating public and worker doses. The expected length of aging will also necessitate some design adjustments. Under 10 CFR 72.236, DCS systems are initially certified for a period of 20 years [2]. Although the Yucca Mountain facility is not intended to be a storage facility under 10 CFR 72, the operational life of the SNF Aging System is 50 years [1]. Any cask system selected for use in aging will have to be qualified to this design lifetime. These considerations are examined, and a summary is provided of the adaptations that must be made in order to use DCS technologies successfully at a geologic repository.« less

Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-04-01

A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15{degrees}C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321{degrees}C. In general, the surface dose ratesmore » were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask.« less

78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

...-0308] RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear... proposing to amend its spent fuel storage regulations by revising the NAC International, Inc., Modular Advanced Generation Nuclear All-purpose Storage (MAGNASTOR[supreg]) Cask System listing within the ``List...

Final design review summary report for the TN-WHC cask and transportation system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kee, A.T.

1997-01-17

This document represents comments generated from a review of Transnuclear`s Final Design Package distributed on December 10, 1996 and a review of the Final Design Analysis Report meeting held on December 17 & 18, 1996. The Final design describes desicn features and presents final analyses @j performed to fabricate and operate the system while meeting the Cask/Transportation Functions and Requirements, WHC-SD-SNF-FRD-011, Rev. 0 and specification WHC-S-0396, Rev. 1.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krementz, Dan; Rose, David; Dunsmuir, Mike

2014-02-06

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

COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 andmore » 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.« less

75 FR 27401 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

... Storage Casks: NUHOMS[reg] HD System Revision 1; Correction AGENCY: Nuclear Regulatory Commission. ACTION... HD spent fuel storage cask system. This action is necessary to correctly specify the effective date... on May 6, 2010 (75 FR 24786), that amends the regulations that govern storage of spent nuclear fuel...

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

High energy neutron transmission analysis of dry cask storage

NASA Astrophysics Data System (ADS)

Greulich, Christopher; Hughes, Christopher; Gao, Yuan; Enqvist, Andreas; Baciak, James

2017-12-01

Since the U.S. currently only approves of storing used nuclear fuel in pools or dry casks, the demand for dry cask storage is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. With the rising demand comes additional pressure to ensure the integrity of dry cask systems. Visual inspection is costly and man-power intensive, so alternative nondestructive testing techniques are desired to insure the continued safe and effective storage of fuel. One such approach being investigated by the University of Florida is neutron based computed tomography. Simulations in MCNP are preformed where D-T energy neutrons are transmitted through the dry cask and measured on the opposite side. If the transmitted signal is clear enough, the interior of the cask can be reconstructed from the measurement of the alterations of neutron signal intensity using standard mathematical techniques developed for medical imaging. Preliminary efforts show a correlation between energy and number of scatters (which is an indication of retention of position information). Work is ongoing to quantify if the correlation is strong enough that an energy discriminator may be used as a filter in future image reconstruction. The calculated transmission probability suggests that an image could be reconstructed with a week of scanning.

Development of a conditioning system for the dual-purpose transport and storage cask for spent nuclear fuel from decommissioned Russian submarines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dyer, R.S.; Barnes, E.; Snipes, R.L.

2007-07-01

Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuelmore » from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to 'Mayak' for reprocessing. The U.S. Environmental Protection Agency (EPA), in cooperation with the U.S. DOD Office of Cooperative Threat Reduction (CTR), and the DOE's ORNL, along with the Norwegian Defense Research Establishment, worked closely with the Ministry of Defense and the Ministry of Atomic Energy of the Russian Federation (RF) to develop an improved integrated management system for interim storage of military SNF in Russia. The initial Project activities included: (1) development of a prototype dual-purpose, metal-concrete 40-ton cask for both the transport and interim storage of RF SNF, and (2) development of the first transshipment/interim storage facility for these casks in Murmansk. The U.S. has continued support to the project by assisting the RF with the development of the first mobile system that provides internal conditioning for the TUK-108/1 casks to allow them to be stored for longer than the current licensing period of two years. Development of the prototype TUK-108/1 cask was completed in December 2000 under the Arctic Military Environmental Cooperation (AMEC) Program. This was the first metal-concrete cask developed, licensed, and produced in the RF for both the transportation and storage of SNF from decommissioned submarines. These casks are currently being serially produced in NW Russia and 108 casks have been produced to date. Russia is using these casks for the transport and interim storage of military SNF from decommissioned nuclear submarines at naval installations in the Arctic and Far East in conformance with the Strategic Arms Reduction Treaty (START II). The design, construction, and commissioning of the first transshipment/interim storage facility in the RF was completed and ready for full operation in September 2003. Because of the RF government reorganization and changing regulations for spent fuel storage facilities, the storage facility at Murmansk was not fully licensed for operation until December 2005. The RF has reported that the facility is now fully operational. The TUK-108/1 SNF transport and storage casks were designed to have a 50-year storage life. Current RF practice is not to condition the submarine SNF or cask during the cask loading. Current RF regulations allow up to 4 mm of residual water (up to 3.2 liters) to remain in the casks. It has been determined that allowing this amount of residual water to remain untreated for a period longer than two years can produce hydrogen gas through hydrolysis which will increase the risk of explosion and could cause some corrosion of internal components. A solution to this problem was to develop and utilize a cask conditioning system to remove the residual water and create an inert storage environment in the cask by back-filling the internal cask cavity with an inert gas, such as helium or argon. This system is compatible with the existing TUK-108/1 design and is mobile for use at multiple submarine dismantlement sites. The RF has required that this cask conditioning system be tested and commissioned at the 'Zvezda' Shipyard in the Far East near Vladivostok, one of the major RF submarine fuel off loading and storage facilities. Currently, the fuel cannot be transferred to 'Mayak' for reprocessing until the completion of the 20 km railroad connector between 'Zvezda' and the main rail line to 'Mayak'. The cask conditioning system will allow extension of the currently-stored casks for an additional three years, at which time the rail connector line should be completed. The current license to store these casks at 'Zvezda' was scheduled to expire on 31 Dec 2006. Without the cask-conditioning system, the license could not be extended, no more fuel could be off-loaded from the decommissioned submarines, and the START objectives could not be met at 'Zvezda'. Completion of this cask conditioning system has removed a significant bottleneck for the completion of the Russian submarine decommissioning program under the START II Agreement. (authors)« less

Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel; Lindgren, Eric R.

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 by increasing the internalmore » 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 aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-08

... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the Holtec HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage... Title 10 of the Code of Federal Regulations Section 72.214 to add the Holtec HI- STORM Flood/Wind cask...

SCAN+

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kenneth Krebs, John Svoboda

2009-11-01

SCAN+ is a software application specifically designed to control the positioning of a gamma spectrometer by a two dimensional translation system above spent fuel bundles located in a sealed spent fuel cask. The gamma spectrometer collects gamma spectrum information for the purpose of spent fuel cask fuel loading verification. SCAN+ performs manual and automatic gamma spectrometer positioning functions as-well-as exercising control of the gamma spectrometer data acquisitioning functions. Cask configuration files are used to determine the positions of spent fuel bundles. Cask scanning files are used to determine the desired scan paths for scanning a spent fuel cask allowing formore » automatic unattended cask scanning that may take several hours.« less

Monte Carlo Shielding Comparative Analysis Applied to TRIGA HEU and LEU Spent Fuel Transport

NASA Astrophysics Data System (ADS)

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

2010-12-01

The paper is a comparative study of LEU and HEU fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for HEU spent fuel, available from the last stage of spent fuel repatriation fulfilled in the summer of 2008, is also presented. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface, and in air at 1 m and 2 m, respectively, from the cask, by means of 3D Monte Carlo MORSE-SGC code. Before loading into the shipping cask, TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. The actinides contribution to total fuel radioactivity is very low in HEU spent fuel case, becoming 10 times greater in LEU spent fuel case. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than HEU ones. Comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, calculated values being greater than the measured ones both to cask wall surface (about 34% relative difference) and in air at 1 m distance from cask surface (about 15% relative difference).

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

PubMed

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. Copyright © 2015 Elsevier Ltd. All rights reserved.

Thermal analyses of the IF-300 shipping cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meier, J.K.

1978-07-01

In order to supply temperature data for structural testing and analysis of shipping casks, a series of thermal analyses using the TRUMP thermal analyzer program were performed on the GE IF-300 spent fuel shipping cask. Major conclusions of the analyses are: (1) Under normal cooling conditions and a cask heat load of 262,000 BTU/h, the seal area of the cask will be roughly 100/sup 0/C (180/sup 0/F) above the ambient surroundings. (2) Under these same conditions the uranium shield at the midpoint of the cask will be between 69/sup 0/C (125/sup 0/F) and 92/sup 0/C (166/sup 0/F) above the ambientmore » surroundings. (3) Significant thermal gradients are not likely to develop between the head studs and the surrounding metal. (4) A representative time constant for the cask as a whole is on the order of one day.« less

75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-16

... Storage Casks: MAGNASTOR System, Revision 1, Confirmation of Effective Date AGENCY: Nuclear Regulatory... spent fuel storage regulations at 10 CFR 72.214 to revise the MAGNASTOR System listing to include...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... has been determined by the NRC. The application must be accompanied by a safety analysis report (SAR). The new SAR may reference the SAR originally submitted for the approved spent fuel storage cask design. (c) The design of a spent fuel storage cask will be reapproved if the conditions in § 72.238 are met...

76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

...-0007] RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY... or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM...: June 13, 2011. SAR Submitted by: Holtec International, Inc. SAR Title: Safety Analysis Report on the HI...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... regulations by revising the Holtec International HI-STORM 100 dry cask storage system listing within the... and safety will be adequately protected. This direct final rule revises the HI-STORM 100 listing in 10...

78 FR 32077 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

... Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct... final rule that would have revised its spent fuel storage regulations to include Amendment No. 3 to... All-purpose Storage (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel...

Array Detector Modules for Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolotnikov, Aleksey

Brookhaven National Laboratory (BNL) proposes to evaluate the arrays of position-sensitive virtual Frisch-grid (VFG) detectors for passive gamma-ray emission tomography (ET) to verify the spent fuel in storage casks before storing them in geo-repositories. Our primary objective is to conduct a preliminary analysis of the arrays capabilities and to perform field measurements to validate the effectiveness of the proposed array modules. The outcome of this proposal will consist of baseline designs for the future ET system which can ultimately be used together with neutrons detectors. This will demonstrate the usage of this technology in spent fuel storage casks.

78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... include: adding a new transfer cask (TC), the OS197L, for use with the 32PT and 61BT dry shielded.... 1004. Specifically, Transnuclear, Inc. requested changes to: (1) add a new TC, the OS197L, for use with... with NUREG-1745 requirements. Deleting the TC dose rates for all currently licensed payloads (TSs 1.2...

Feasibility study for a transportation operations system cask maintenance facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 caskmore » 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.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

... Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation Nuclear All-purpose Storage...

10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

Code of Federal Regulations, 2013 CFR

2013-01-01

... to exceed 40 years. In the event that the certificate holder does not apply for a cask design renewal...) The application must be accompanied by a safety analysis report (SAR). The SAR must include the following: (1) Design bases information as documented in the most recently updated final safety analysis...

10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

Code of Federal Regulations, 2012 CFR

2012-01-01

... to exceed 40 years. In the event that the certificate holder does not apply for a cask design renewal...) The application must be accompanied by a safety analysis report (SAR). The SAR must include the following: (1) Design bases information as documented in the most recently updated final safety analysis...

10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

Code of Federal Regulations, 2014 CFR

2014-01-01

... to exceed 40 years. In the event that the certificate holder does not apply for a cask design renewal...) The application must be accompanied by a safety analysis report (SAR). The SAR must include the following: (1) Design bases information as documented in the most recently updated final safety analysis...

Radioactive materials shipping cask anticontamination enclosure

DOEpatents

Belmonte, Mark S.; Davis, James H.; Williams, David A.

1982-01-01

An anticontamination device for use in storing shipping casks for radioactive materials comprising (1) a seal plate assembly; (2) a double-layer plastic bag; and (3) a water management system or means for water management.

Evaluation of Cask Drop Criticality Issues at K Basin

DOE Office of Scientific and Technical Information (OSTI.GOV)

GOLDMANN, L.H.

An analysis of ability of Multi-canister Overpack (MCO) to withstand drops at K Basin without exceeding the criticality design requirements. Report concludes the MCO will function acceptably. The spent fuel currently residing in the 105 KE and 105 KW storage basins will be placed in fuel storage baskets which will be loaded into the MCO cask assembly. During the basket loading operations the MCO cask assembly will be positioned near the bottom of the south load out pit (SLOP). The loaded MCO cask will be lifted from the SLOP transferred to the transport trailer and delivered to the Cold Vacuummore » Drying Facility (CVDF). In the wet condition there is a potential for criticality problems if significant changes in the designed fuel configurations occur. The purpose of this report is to address structural issues associated with criticality design features for MCO cask drop accidents in the 105 KE and 105 KW facilities.« less

75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... modifications to the Vertical Concrete Cask (VCC) incorporating design features from the MAGNASTOR system for...; an increase in the concrete pad compression strength from 4,000 psi to 6,000 psi; added justification... system while adhering to ALARA principles; (5) an increase in the concrete pad compression strength from...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 inmore » 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.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-25

... Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... revising the Holtec International HI-STORM 100 System listing within the ``List of Approved Spent Fuel...) 72.214, by revising the Holtec International HI-STORM 100 System listing within the ``List of...

75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

...; minor design modifications to the Vertical Concrete Cask (VCC) incorporating design features from the... (ALARA) principles; an increase in the concrete pad compression strength from 4000 psi to 6000 psi; added...

Storage, transportation and disposal system for used nuclear fuel assemblies

DOEpatents

Scaglione, John M.; Wagner, John C.

2017-01-10

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

Spent nuclear fuel dry transfer system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, L.; Agace, S.

The U.S. Department of Energy is currently engaged in a cooperative program with the Electric Power Research Institute (EPRI) to design a spent nuclear fuel dry transfer system (DTS). The system will enable the transfer of individual spent nuclear fuel assemblies between a conventional top loading cask and multi-purpose canister in a shielded overpack, or accommodate spent nuclear fuel transfers between two conventional casks.

78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-27

... dangerous to living organisms, including insects, microbes, bacteria or virus that attach to dust that.... According to his comment, ``[T]he Deer Tick has carried a spirochete bacteria for millions of years, but...

78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

...: Gregory R. Trussell, Office of Federal and State Materials and Environmental Management Programs, U.S... Access and Management System (ADAMS): You may access publicly-available documents online in the NRC... continues to be ensured. The direct final rule will become effective on January 7, 2014. However, if the NRC...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryan, Charles R.

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. Thismore » 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.« less

Characterization of neutron sources from spent fuel casks. [Skyshine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parks, C.V.; Pace, J.V. III

1987-01-01

In the interim period prior to the acceptance of spent fuel for disposal by the USDOE, utilities are beginning to choose dry cask storage as an alternative to pool re-racking, transshipments, or new pool construction. In addition, the current MRS proposal calls for interim dry storage of consolidated spent fuel in concrete casks. As part of the licensing requirements for these cask storage facilities, calculations are typically necessary to determine the yearly radiation dose received at the site boundary. Unlike wet facilities, neutron skyshine can be an important contribution to the total boundary dose from a dry storage facility. Calculationmore » of the neutron skyshine is in turn heavily dependent on the source characteristics and source model selected for the analysis. This paper presents the basic source characteristics of the spent fuel stored in dry casks and discusses factors that must be considered in evaluating and modeling the radiation sources for the subsequent skyshine calculation. 4 refs., 1 tab.« less

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

PubMed

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-B 4 C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-25

... Specification (TS) Surveillance Requirement 3.1.6.1 to verify the operability of the concrete cask heat removal....6.1 to verify the operability of the concrete cask heat removal system to maintain safe storage...

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.

PubMed

Huang, Tzyy-Nan; Hsueh, Yi-Ping

2017-01-01

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. 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. 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. The functions of other CASK protein interactions cannot be addressed using CASK T740A mice. Disruption of the CASK and TBR1 interaction impairs extinction, suggesting the involvement of CASK in cognitive flexibility.

NEUTRON CHARACTERIZATION OF ENSA-DPT TYPE SPENT FUEL CASK AT TRILLO NUCLEAR POWER PLANT.

PubMed

Méndez-Villafañe, Roberto; Campo-Blanco, Xandra; Embid, Miguel; Yéboles, César A; Morales, Ramón; Novo, Manuel; Sanz, Javier

2018-04-23

The Neutron Standards Laboratory of CIEMAT has conducted the characterization of the independent spent fuel storage installation at the Trillo Nuclear Power Plant. At this facility, the spent fuel assemblies are stored in ENSA-DPT type dual purpose casks. Neutron characterization was performed by dosimetry measurements with a neutron survey meter (LB6411) inside the facility, around an individual cask and between stored casks, and outside the facility. Spectra measurements were also performed with a Bonner sphere system in order to determine the integral quantities and validate the use of the neutron monitor at the different positions. Inside the facility, measured neutron spectra and neutron ambient dose equivalent rate are consistent with the casks spatial distribution and neutron emission rates, and measurements with both instruments are consistent with each other. Outside the facility, measured neutron ambient dose equivalent rates are well below the 0.5 μSv/h limit established by the nuclear regulatory authority.

75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-23

... Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective Date AGENCY: Nuclear... include Amendment Number 6 to Certificate of Compliance (CoC) Number 1025. DATES: Effective Date: The... regulations at 10 CFR 72.214 to include Amendment No. 6 to CoC No. 1025. Amendment No. 6 changes the...

Storage, transportation and disposal system for used nuclear fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M.; Wagner, John C.

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

Development of Friction Stir Processing for Repair of Nuclear Dry Cask Storage System Canisters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Kenneth A.; Sutton, Ben; Grant, Glenn J.

The Nuclear Regulatory Commission has identified chloride-induced stress corrosion cracking (CISCC) of austenitic stainless steel dry cask storage systems (DCSS) as an area of great concern. Friction Stir Processing (FSP) was used to repair laboratory-generated stress corrosion cracking (SCC) in representative stainless steel 304 coupons. Results of this study show FSP is a viable method for repair and mitigation CISCC. This paper highlights lessons learned and developed techniques relative to FSP development for crack repair in sensitized thick section stainless steel 304. These include: development of process parameters, welding at low spindle speed, use of weld power and temperature controlmore » and optimization of these controls. NDE and destructive analysis are also presented to demonstrate effectiveness of the developed methods for SCC crack repair.« less

Smart Winery: A Real-Time Monitoring System for Structural Health and Ullage in Fino Style Wine Casks

PubMed Central

Cañete, Eduardo; Chen, Jaime; Rubio, Bartolomé

2018-01-01

The rapid development in low-cost sensor and wireless communication technology has made it possible for a large number of devices to coexist and exchange information autonomously. It has been predicted that a substantial number of devices will be able to exchange and provide information about an environment with the goal of improving our lives, under the well-known paradigm of the Internet of Things (IoT). One of the main applications of these kinds of devices is the monitoring of scenarios. In order to improve the current wine elaboration process, this paper presents a real-time monitoring system to supervise the status of wine casks. We have focused on a special kind of white wine, called Fino, principally produced in Andalusia (Southern Spain). The process by which this kind of wind is monitored is completely different from that of red wine, as the casks are not completely full and, due to the fact that they are not renewed very often, are more prone to breakage. A smart cork prototype monitors the structural health, the ullage, and the level of light inside the cask and the room temperature. The advantage of this smart cork is that it allows winemakers to monitor, in real time, the status of each wine cask so that, if an issue is detected (e.g., a crack appears in the cask), they can act immediately to resolve it. Moreover, abnormal parameters or incorrect environmental conditions can be detected in time before the wine loses its desired qualities. The system has been tested in “Bodegas San Acacio,” a winery based in Montemayor, a town in the north of Andalusia. Results show that the use of such a system can provide a solution that tracks the evolution and assesses the suitability of the delicate wine elaboration process in real time, which is especially important for the kind of wine considered in this paper. PMID:29518928

Smart Winery: A Real-Time Monitoring System for Structural Health and Ullage in Fino Style Wine Casks.

PubMed

Cañete, Eduardo; Chen, Jaime; Martín, Cristian; Rubio, Bartolomé

2018-03-07

The rapid development in low-cost sensor and wireless communication technology has made it possible for a large number of devices to coexist and exchange information autonomously. It has been predicted that a substantial number of devices will be able to exchange and provide information about an environment with the goal of improving our lives, under the well-known paradigm of the Internet of Things (IoT). One of the main applications of these kinds of devices is the monitoring of scenarios. In order to improve the current wine elaboration process, this paper presents a real-time monitoring system to supervise the status of wine casks. We have focused on a special kind of white wine, called Fino, principally produced in Andalusia (Southern Spain). The process by which this kind of wind is monitored is completely different from that of red wine, as the casks are not completely full and, due to the fact that they are not renewed very often, are more prone to breakage. A smart cork prototype monitors the structural health, the ullage, and the level of light inside the cask and the room temperature. The advantage of this smart cork is that it allows winemakers to monitor, in real time, the status of each wine cask so that, if an issue is detected (e.g., a crack appears in the cask), they can act immediately to resolve it. Moreover, abnormal parameters or incorrect environmental conditions can be detected in time before the wine loses its desired qualities. The system has been tested in "Bodegas San Acacio," a winery based in Montemayor, a town in the north of Andalusia. Results show that the use of such a system can provide a solution that tracks the evolution and assesses the suitability of the delicate wine elaboration process in real time, which is especially important for the kind of wine considered in this paper.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 themore » 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.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-23

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0250] Computational Fluid Dynamics Best Practice... public comments on draft NUREG-2152, ``Computational Fluid Dynamics Best Practice Guidelines for Dry Cask... System (ADAMS): You may access publicly-available documents online in the NRC Library at http://www.nrc...

A review of ventilated storage cask (VSC) system projects and experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConaghy, W.

1995-12-31

First, the author discusses the ventilated storage cask (VSC) design and an operations summary is given. Next VSC project status at Palisades, Point Beach, Arkansas Nuclear One, Fast Flux Test Facility and Zaporozhye is discussed. Lastly, VSC operational experience and VSC transportation interfaces are reviewed.

CASK regulates CaMKII autophosphorylation in neuronal growth, calcium signaling, and learning

PubMed Central

Gillespie, John M.; Hodge, James J. L.

2013-01-01

Calcium (Ca2+)/calmodulin (CaM)-dependent kinase II (CaMKII) activity plays a fundamental role in learning and memory. A key feature of CaMKII in memory formation is its ability to be regulated by autophosphorylation, which switches its activity on and off during synaptic plasticity. The synaptic scaffolding protein CASK (calcium (Ca2+)/calmodulin (CaM) associated serine kinase) is also important for learning and memory, as mutations in CASK result in intellectual disability and neurological defects in humans. We show that in Drosophila larvae, CASK interacts with CaMKII to control neuronal growth and calcium signaling. Furthermore, deletion of the CaMK-like and L27 domains of CASK (CASK β null) or expression of overactive CaMKII (T287D) produced similar effects on synaptic growth and Ca2+ signaling. CASK overexpression rescues the effects of CaMKII overactivity, consistent with the notion that CASK and CaMKII act in a common pathway that controls these neuronal processes. The reduction in Ca2+ signaling observed in the CASK β null mutant caused a decrease in vesicle trafficking at synapses. In addition, the decrease in Ca2+ signaling in CASK mutants was associated with an increase in Ether-à-go-go (EAG) potassium (K+) channel localization to synapses. Reducing EAG restored the decrease in Ca2+ signaling observed in CASK mutants to the level of wildtype, suggesting that CASK regulates Ca2+ signaling via EAG. CASK knockdown reduced both appetitive associative learning and odor evoked Ca2+ responses in Drosophila mushroom bodies, which are the learning centers of Drosophila. Expression of human CASK in Drosophila rescued the effect of CASK deletion on the activity state of CaMKII, suggesting that human CASK may also regulate CaMKII autophosphorylation. PMID:24062638

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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,more » 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.« less

Thermal evaluation of alternative shipping cask for irradiated experiments

DOE PAGES

Guillen, Donna Post

2015-06-01

Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. 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 PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavitymore » of the GE-2000 cask is too short to accommodate shipping the larger payloads. 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 payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.« less

Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dilger, Fred; Halstead, Robert J.; Ballard, James D.

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very largemore » dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Sanghoon; Choi, Woo-Seok; Kim, Ki-Young

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 themore » 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)« less

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa

2006-04-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstrationmore » project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

..., and criticality control. If there is no loss of confinement, shielding, or criticality control, the... would prevent loss of confinement, shielding, and criticality control. If there is no loss of...;Federal Register / Vol. 78, No. 235 / Friday, December 6, 2013 / Rules and Regulations#0;#0; [[Page 73379...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Safety analysis report for packaging (onsite) multicanister overpack cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Proposed... spent fuel storage cask regulations by revising the Holtec International HI-STORM 100 dry cask storage... Amendment No. 8 to CoC No. 1014 and does not include other aspects of the HI-STORM 100 dry storage cask...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 anymore » 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.« less

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Ross, Steven; Grey, Carissa Ann

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacificmore » Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ibarra, Luis; Sanders, David; Yang, Haori

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 tomore » 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 may also slide and collide with other casks or structural components. Also, the different DSC components may impact each other during these events. This study provides a comprehensive evaluation of DSCs subjected to these extreme demands, including the effect of vertical accelerations, and soilstructure interaction.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, K.; Bellamy, S.; Daugherty, W.

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

Characterization of the radiation environment for a large-area interim spent-nuclear-fuel storage facility

NASA Astrophysics Data System (ADS)

Fortkamp, Jonathan C.

Current needs in the nuclear industry and movements in the political arena indicate that authorization may soon be given for development of a federal interim storage facility for spent nuclear fuel. The initial stages of the design work have already begun within the Department of Energy and are being reviewed by the Nuclear Regulatory Commission. This dissertation addresses the radiation environment around an interim spent nuclear fuel storage facility. Specifically the dissertation characterizes the radiation dose rates around the facility based on a design basis source term, evaluates the changes in dose due to varying cask spacing configurations, and uses these results to define some applicable health physics principles for the storage facility. Results indicate that dose rates from the facility are due primarily from photons from the spent fuel and Co-60 activation in the fuel assemblies. In the modeled cask system, skyshine was a significant contribution to dose rates at distances from the cask array, but this contribution can be reduced with an alternate cask venting system. With the application of appropriate health physics principles, occupation doses can be easily maintained far below regulatory limits and maintained ALARA.

Multiple-Angle Muon Radiography of a Dry Storage Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jianmin; Bazant, Zdenek; Jacobs, Laurence

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 ismore » 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« less

Used Fuel Cask Identification through Neutron Profile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

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.more » 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.« less

TN International and ITS operational feedback regarding the decommissioning of obsolete casks dedicated to the transport and/or storage of nuclear raw materials, fuel and used fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blachet, L.; Bimet, F.; Rennesson, N.

2008-07-01

Within the AREVA group, TN International is a major actor regarding the design of casks and transportation for the nuclear cycle. In the early 2005, TN International has started the project of decommissioning some of its own equipment and was hence the first company ever in the AREVA Group to implement this new approach. In order to do so, TN International has based this project by taking into account the AREVA Sustainable Development Charter, the French regulatory framework, the ANDRA (Agence Nationale pour la Gestion des Dechets Radioactifs - National Agency for the radioactive waste management) requirements and has deployedmore » a step by step methodology such as radiological characterization following a logical route. The aim was to define a standardized process with optimized solutions regarding the diversity of the cask's fleet. As a general matter, decommissioning of nuclear casks is a brand new field as the nuclear field is more familiar with the dismantling of nuclear facilities and/or nuclear power plant. Nevertheless existing workshops, maintenance facilities, measurements equipments and techniques have been exploited and adapted by TN International in order to turn an ambitious project into a permanent and cost-effective activity. The decommissioning of the nuclear casks implemented by TN International regarding its own needs and the French regulatory framework is formalized by several processes and is materialized for instance by the final disposal of casks as they are or in ISO container packed with cut-off casks and big bags filled with crushed internal cask equipments, etc. The first part of this paper aims to describe the history of the project that started with a specific environmental analysis which took into account the values of AREVA as regards the Sustainable Development principles that were at the time and are still a topic of current concern in the world. The second part will deal with the definition, the design and the implementation of the decommissioning processes and the applied techniques. The third part will present a two years operational feedback. The last part will introduce new processes which are currently under investigation and will put into light that decommissioning of nuclear casks is a continuous activity that is in perpetual mutation. (authors)« less

Measurement of chlorine concentration on steel surfaces via fiber-optic laser-induced breakdown spectroscopy in double-pulse configuration

NASA Astrophysics Data System (ADS)

Xiao, X.; Le Berre, S.; Fobar, D. G.; Burger, M.; Skrodzki, P. J.; Hartig, K. C.; Motta, A. T.; Jovanovic, I.

2018-03-01

The corrosive environment provided by chlorine ions on the welds of stainless steel dry cask storage canisters for used nuclear fuel may contribute to the occurrence of stress corrosion cracking. We demonstrate the use of fiber-optic laser-induced breakdown spectroscopy (FOLIBS) in the double-pulse (DP) configuration for high-sensitivity, remote measurement of the surface concentrations of chlorine compatible in constrained space and challenging environment characteristic for dry cask storage systems. Chlorine surface concentrations as low as 5 mg/m2 have been detected and quantified by use of a laboratory-based and a fieldable DP FOLIBS setup with the calibration curve approach. The compact final optics assembly in the fieldable setup is interfaced via two 25-m long optical fibers for high-power laser pulse delivery and plasma emission collection and can be readily integrated into a multi-sensor robotic delivery system for in-situ inspection of dry cask storage systems.

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

PubMed Central

Dembowski, Jill A.; An, Ping; Scoulos-Hanson, Maritsa; Yeo, Gene; Han, Joonhee; Fu, Xiang-Dong; Grabowski, Paula J.

2012-01-01

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. PMID:23008758

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 tomore » 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.« less

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Jie; Liu, Yung Y.

2016-05-01

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 x 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. Themore » effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask.« less

Adsorbed radioactivity and radiographic imaging of surfaces of stainless steel and titanium

NASA Astrophysics Data System (ADS)

Jung, Haijo

1997-11-01

Type 304 stainless steel used for typical surface materials of spent fuel shipping casks and titanium were exposed in the spent fuel storage pool of a typical PWR power plant. Adsorption characteristics, effectiveness of decontamination by water cleaning and by electrocleaning, and swipe effectiveness on the metal surfaces were studied. A variety of environmental conditions had been manipulated to stimulate the potential 'weeping' phenomenon that often occurs with spent fuel shipping casks during transit. In a previous study, few heterogeneous effects of adsorbed contamination onto metal surfaces were observed. Radiographic images of cask surfaces were made in this study and showed clearly heterogeneous activity distributions. Acquired radiographic images were digitized and further analyzed with an image analysis computer package and compared to calibrated images by using standard sources. The measurements of activity distribution by using the radiographic image method were consistent with that using a HPGe detector. This radiographic image method was used to study the effects of electrocleaning for total and specified areas. The Modulation Transfer Function (MTF) of a film-screen system in contact with a radioactive metal surface was studied with neutron activated gold foils and showed more broad resolution properties than general diagnostic x-ray film-screen systems. Microstructure between normal areas and hot spots showed significant differences, and one hot spot appearing as a dot on the film image consisted of several small hot spots (about 10 μm in diameter). These hot spots were observed as structural defects of the metal surfaces.

Nondestructive Evaluation of the VSC-17 Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl Morton; Al Carlson; Cecilia Hoffman

2006-01-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC 17) spent nuclear fuel storage cask, originally located at the INL Test Area North, as a candidate to study cask performance because it had been used to storemore » fuel as part of a dry cask storage demonstration project for over 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. The INL team met with the CRIEPI representatives in December of 2004 to discuss the next steps. As a result of that meeting, CRIEPI requested that in the summer 2005 INL perform additional surveys on the VSC 17 cask with participation of CRIEPI scientists. This document summarizes the evaluation methods used on the VSC 17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bracey, William; Bondre, Jayant; Shelton, Catherine

2013-07-01

The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In Januarymore » 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O.K.; Diercks, D.; Fabian, R.

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2014-09-01

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

Early thermal testing of type B radioactive material packages in USA to environments beyond regulatory package thermal test standards

DOE PAGES

Yoshimura, H. R.; Pope, R. B.; Kubo, M.

2007-06-01

Three separate fire test programmes exposing casks beyond the regulatory thermal test requirements were performed by Sandia National Laboratories during the late 1970s and mid 1980s. The results of these test programmes can be used to assist in addressing the adequacy of the regulatory thermal test of fully engulfing exposure at 800°C for 30 min and how that test might relate to real accident thermal environments. The test programmes were undertaken on obsolete and new casks on behalf of the US Department of Energy (DOE), the US Department of Transportation (DOT) and the Japanese Power Reactor and Nuclear Fuel Developmentmore » Corporation (PNC), currently known as the Japan Atomic Energy Agency. Two of the tests involved exposure of casks in damaged transport vehicles to fully engulfing fires for 72–125 min, and the other test involved four exposures of a cask to torch environments for 30 min. Much of the original documentation regarding these tests and their results is no longer readily available. The documents relating to these tests have been surveyed; this paper presents summaries from this survey of the tests and their results. Specifically, for the pool fire exposures, the temperatures measured in the flames of both exceeded the flame temperature required by the Transport Regulations; yet an obsolete 67 t cask endured 90 min of exposure before evidence of failure was detected, and a new cask endured the 72 min exposure while retaining its containment integrity. For the exposure of a modified obsolete cask to four different torch environments, the integrity of the cask was retained and the relative temperature increases within the cask were well within acceptable limits and well below the values that could be expected if the cask was exposed to the regulatory thermal test. In this paper, a review of these three thermal test programmes, establishes that the two older cask designs and one new cask design have the ability to survive environments that were different from (the torch environments) or more severe than the environment specified by the existing thermal test requirement in the Transport Regulations. Finally, these results can be extrapolated to apply to modern casks that generally have more robust designs as well as better quality assurance applied during the manufacturing process.« less

CASK and CaMKII function in Drosophila memory

PubMed Central

Malik, Bilal R.; Hodge, James J. L.

2014-01-01

Calcium (Ca2+) and Calmodulin (CaM)-dependent serine/threonine kinase II (CaMKII) plays a central role in synaptic plasticity and memory due to its ability to phosphorylate itself and regulate its own kinase activity. Autophosphorylation at threonine 287 (T287) switches CaMKII to a Ca2+ independent and constitutively active state replicated by overexpression of a phosphomimetic CaMKII-T287D transgene or blocked by expression of a T287A transgene. A second pair of sites, T306 T307 in the CaM binding region once autophosphorylated, prevents CaM binding and inactivates the kinase during synaptic plasticity and memory, and can be blocked by a TT306/7AA transgene. Recently the synaptic scaffolding molecule called CASK (Ca2+/CaM-associated serine kinase) has been shown to control both sets of CaMKII autophosphorylation events during neuronal growth, Ca2+ signaling and memory in Drosophila. Deletion of either full length CASK or just its CaMK-like and L27 domains removed middle-term memory (MTM) and long-term memory (LTM), with CASK function in the α′/ß′ mushroom body neurons being required for memory. In a similar manner directly changing the levels of CaMKII autophosphorylation (T287D, T287A, or TT306/7AA) in the α′/ß′ neurons also removed MTM and LTM. In the CASK null mutant expression of either the Drosophila or human CASK transgene in the α′/ß′ neurons was found to completely rescue memory, confirming that CASK signaling in α′/β′ neurons is necessary and sufficient for Drosophila memory formation and that the neuronal function of CASK is conserved between Drosophila and human. Expression of human CASK in Drosophila also rescued the effect of CASK deletion on the activity state of CaMKII, suggesting that human CASK may also regulate CaMKII autophosphorylation. Mutations in human CASK have recently been shown to result in intellectual disability and neurological defects suggesting a role in plasticity and learning possibly via regulation of CaMKII autophosphorylation. PMID:25009461

CASK and CaMKII function in the mushroom body α'/β' neurons during Drosophila memory formation.

PubMed

Malik, Bilal R; Gillespie, John Michael; Hodge, James J L

2013-01-01

Ca(2+)/CaM serine/threonine kinase II (CaMKII) is a central molecule in mechanisms of synaptic plasticity and memory. A vital feature of CaMKII in plasticity is its ability to switch to a calcium (Ca(2+)) independent constitutively active state after autophosphorylation at threonine 287 (T287). A second pair of sites, T306 T307 in the calmodulin (CaM) binding region once autophosphorylated, prevent subsequent CaM binding and inactivates the kinase during synaptic plasticity and memory. Recently a synaptic molecule called Ca(2+)/CaM-dependent serine protein kinase (CASK) has been shown to control both sets of CaMKII autophosphorylation events and hence is well poised to be a key regulator of memory. We show deletion of full length CASK or just its CaMK-like and L27 domains disrupts middle-term memory (MTM) and long-term memory (LTM), with CASK function in the α'/β' subset of mushroom body neurons being required for memory. Likewise directly changing the levels of CaMKII autophosphorylation in these neurons removed MTM and LTM. The requirement of CASK and CaMKII autophosphorylation was not developmental as their manipulation just in the adult α'/β' neurons was sufficient to remove memory. Overexpression of CASK or CaMKII in the α'/β' neurons also occluded MTM and LTM. Overexpression of either Drosophila or human CASK in the α'/β' neurons of the CASK mutant completely rescued memory, confirming that CASK signaling in α'/β' neurons is necessary and sufficient for Drosophila memory formation and that the neuronal function of CASK is conserved between Drosophila and human. At the cellular level CaMKII overexpression in the α'/β' neurons increased activity dependent Ca(2+) responses while reduction of CaMKII decreased it. Likewise reducing CASK or directly expressing a phosphomimetic CaMKII T287D transgene in the α'/β' similarly decreased Ca(2+) signaling. Our results are consistent with CASK regulating CaMKII autophosphorylation in a pathway required for memory formation that involves activity dependent changes in Ca(2+) signaling in the α'/β' neurons.

Viability of Existing INL Facilities for Dry Storage Cask Handling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Randy Bohachek; Charles Park; Bruce Wallace

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 Hotmore » 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.« less

Viability of Existing INL Facilities for Dry Storage Cask Handling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohachek, Randy; Wallace, Bruce; Winston, Phil

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 Hotmore » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 bemore » 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.« less

Genetics Home Reference: CASK-related intellectual disability

MedlinePlus

... XL-ID with or without nystagmus (rapid, involuntary eye movements) is a milder form of CASK -related intellectual ... to promote development of the nerves that control eye movement (the oculomotor neural network). Mutations in the CASK ...

Computational Fluid Dynamics Best Practice Guidelines in the Analysis of Storage Dry Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zigh, A.; Solis, J.

2008-07-01

Computational fluid dynamics (CFD) methods are used to evaluate the thermal performance of a dry cask under long term storage conditions in accordance with NUREG-1536 [NUREG-1536, 1997]. A three-dimensional CFD model was developed and validated using data for a ventilated storage cask (VSC-17) collected by Idaho National Laboratory (INL). The developed Fluent CFD model was validated to minimize the modeling and application uncertainties. To address modeling uncertainties, the paper focused on turbulence modeling of buoyancy driven air flow. Similarly, in the application uncertainties, the pressure boundary conditions used to model the air inlet and outlet vents were investigated and validated.more » Different turbulence models were used to reduce the modeling uncertainty in the CFD simulation of the air flow through the annular gap between the overpack and the multi-assembly sealed basket (MSB). Among the chosen turbulence models, the validation showed that the low Reynolds k-{epsilon} and the transitional k-{omega} turbulence models predicted the measured temperatures closely. To assess the impact of pressure boundary conditions used at the air inlet and outlet channels on the application uncertainties, a sensitivity analysis of operating density was undertaken. For convergence purposes, all available commercial CFD codes include the operating density in the pressure gradient term of the momentum equation. The validation showed that the correct operating density corresponds to the density evaluated at the air inlet condition of pressure and temperature. Next, the validated CFD method was used to predict the thermal performance of an existing dry cask storage system. The evaluation uses two distinct models: a three-dimensional and an axisymmetrical representation of the cask. In the 3-D model, porous media was used to model only the volume occupied by the rodded region that is surrounded by the BWR channel box. In the axisymmetric model, porous media was used to model the entire region that encompasses the fuel assemblies as well as the gaps in between. Consequently, a larger volume is represented by porous media in the second model; hence, a higher frictional flow resistance is introduced in the momentum equations. The conservatism and the safety margins of these models were compared to assess the applicability and the realism of these two models. The three-dimensional model included fewer geometry simplifications and is recommended as it predicted less conservative fuel cladding temperature values, while still assuring the existence of adequate safety margins. (authors)« less

Test Plan for Cask Identification Detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

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.

CASK and CaMKII function in the mushroom body α′/β′ neurons during Drosophila memory formation

PubMed Central

Malik, Bilal R.; Gillespie, John Michael; Hodge, James J. L.

2013-01-01

Ca2+/CaM serine/threonine kinase II (CaMKII) is a central molecule in mechanisms of synaptic plasticity and memory. A vital feature of CaMKII in plasticity is its ability to switch to a calcium (Ca2+) independent constitutively active state after autophosphorylation at threonine 287 (T287). A second pair of sites, T306 T307 in the calmodulin (CaM) binding region once autophosphorylated, prevent subsequent CaM binding and inactivates the kinase during synaptic plasticity and memory. Recently a synaptic molecule called Ca2+/CaM-dependent serine protein kinase (CASK) has been shown to control both sets of CaMKII autophosphorylation events and hence is well poised to be a key regulator of memory. We show deletion of full length CASK or just its CaMK-like and L27 domains disrupts middle-term memory (MTM) and long-term memory (LTM), with CASK function in the α′/β′ subset of mushroom body neurons being required for memory. Likewise directly changing the levels of CaMKII autophosphorylation in these neurons removed MTM and LTM. The requirement of CASK and CaMKII autophosphorylation was not developmental as their manipulation just in the adult α′/β′ neurons was sufficient to remove memory. Overexpression of CASK or CaMKII in the α′/β′ neurons also occluded MTM and LTM. Overexpression of either Drosophila or human CASK in the α′/β′ neurons of the CASK mutant completely rescued memory, confirming that CASK signaling in α′/β′ neurons is necessary and sufficient for Drosophila memory formation and that the neuronal function of CASK is conserved between Drosophila and human. At the cellular level CaMKII overexpression in the α′/β′ neurons increased activity dependent Ca2+ responses while reduction of CaMKII decreased it. Likewise reducing CASK or directly expressing a phosphomimetic CaMKII T287D transgene in the α′/β′ similarly decreased Ca2+ signaling. Our results are consistent with CASK regulating CaMKII autophosphorylation in a pathway required for memory formation that involves activity dependent changes in Ca2+ signaling in the α′/β′ neurons. PMID:23543616

Signatures of Extended Storage of Used Nuclear Fuel in Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

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 Controlmore » 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.« less

Calcium/calmodulin-dependent serine protein kinase CASK modulates the L-type calcium current.

PubMed

Nafzger, Sabine; Rougier, Jean-Sebastien

2017-01-01

The L-type voltage-gated calcium channel Ca v 1.2 mediates the calcium influx into cells upon membrane depolarization. The list of cardiopathies associated to Ca v 1.2 dysfunctions highlights the importance of this channel in cardiac physiology. Calcium/calmodulin-dependent serine protein kinase (CASK), expressed in cardiac cells, has been identified as a regulator of Ca v 2.2 channels in neurons, but no experiments have been performed to investigate its role in Ca v 1.2 regulation. Full length or the distal C-terminal truncated of the pore-forming Ca v 1.2 channel (Ca v 1.2α1c), both present in cardiac cells, were expressed in TsA-201 cells. In addition, a shRNA silencer, or scramble as negative control, of CASK was co-transfected in order to silence CASK endogenously expressed. Three days post-transfection, the barium current was increased only for the truncated form without alteration of the steady state activation and inactivation biophysical properties. The calcium current, however, was increased after CASK silencing with both types of Ca v 1.2α1c subunits suggesting that, in absence of calcium, the distal C-terminal counteracts the CASK effect. Biochemistry experiments did not reveals neither an alteration of Ca v 1.2 channel protein expression after CASK silencing nor an interaction between Ca v 1.2α1c subunits and CASK. Nevertheless, after CASK silencing, single calcium channel recordings have shown an increase of the voltage-gated calcium channel Ca v 1.2 open probability explaining the increase of the whole-cell current. This study suggests CASK as a novel regulator of Ca v 1.2 via a modulation of the voltage-gated calcium channel Ca v 1.2 open probability. Copyright © 2016 Elsevier Ltd. All rights reserved.

CASK interacts with PMCA4b and JAM-A on the Mouse Sperm Flagellum to Regulate Ca2+ Homeostasis and Motility1

PubMed Central

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.

2012-01-01

Deletion of the highly conserved gene for the major Ca2+ 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<0.001) ATP levels, significantly (P<0.001) greater cytosolic Ca2+ concentration ([Ca2+]c) and ~10-fold higher mitochondrial sequestration, indicating Ca2+ overload. Investigating the mechanism involved, we used coimmunoprecipitation studies to show that CASK (Ca2+/calmodulin-dependent serine kinase), identified for the first time on the sperm flagellum where it co-localizes with both PMCA4b and 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 Ca2+ 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 Ca2+ homeostasis in sperm is maintained by the relative ratios of CASK-PMCA4b and CASK-JAM-A interactions. PMID:22020416

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

DTIC Science & Technology

1983-06-01

Axial Reference Point ( XRP ) .......... 19 4. A View of the Torch Facility with the Nozzle Directed Side-On to the HNPF Cask... XRP and the TIC for Various HNPF Cask Surfaces in Test Number 1 .................... 47 16. The Spatial Distribution of Sensors in a Cross-Sectional...Plane Through the HNPF Cask at 289.6 cm from the XRP as Viewed from the Top End with the TIC Located at 900 for Test Numbers 1 and 2

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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, withmore » 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.« less

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

NASA Astrophysics Data System (ADS)

Durham, J. M.; Poulson, D.; Bacon, J.; Chichester, D. L.; Guardincerri, E.; Morris, C. L.; Plaud-Ramos, K.; Schwendiman, W.; Tolman, J. D.; Winston, P.

2018-04-01

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. Here we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. This application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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 scattering 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, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release frommore » the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

InstrumentationPod (IPOD) User Guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Robert F.

2014-06-12

This document describes the Instrumentation Pod (IPOD) and its operation and use. The IPOD is a low-power detector system comprising a 3He tube with preamp for neutron detection, a microcontroller-based data acquisition system, a GPS receiver for locationdetermination and time-synchronization, and power filtering and protection. The IPOD is intended to be bolted to the top of Dual-Use Casks stored at Baikal-1 in Kazakhstan in order to maintain continuity of knowledge of the materials stored within the cask. The data acquisition system receives pulses from the neutron-detection preamp, combines this information with other sensor data, and stores the result on twomore » SD cards that are part of the data acquisition system. Firmware in the data acquisition system controls collection and storing of the data and enables configuration of the acquisition parameters.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Bisset

This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage areamore » of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known.« less

Canister Design for Deep Borehole Disposal of Nuclear Waste

DTIC Science & Technology

2006-05-01

radioactive waste disposal (not yet released) Fortunately, transportation casks for spent fuel have already been approved, built, and used as...would allow use of the current designs for transportation casks ; or, place the fuel assemblies into the final disposal canisters 21 prior to transport ...16 Figure 1-5. Typical Spent Fuel Transportation Casks

Systems and methods for harvesting and storing materials produced in a nuclear reactor

DOEpatents

Heinold, Mark R.; Dayal, Yogeshwar; Brittingham, Martin W.

2016-04-05

Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luna, R. E.

This paper provides a simple model for estimating the release of respirable aerosols resulting from an attack on a spent fuel cask using a high energy density device (HEDD). Two primary experiments have provided data on potential releases from spent fuel casks under HEDD attack. Sandia National Laboratories (SNL) conducted the first in the early 1980's and the second was sponsored by Gessellshaft fur Anlagen- and Reaktorsicherheit (GRS) in Germany and conducted in France in 1994. Both used surrogate spent fuel assemblies in real casks. The SNL experiments used un-pressurized fuel pin assemblies in a single element cask while themore » GRS tests used pressurized fuel pin assemblies in a 9-element cask. Data from the two test programs is reasonably consistent, given the differences in the experiments, but the use of the test data for prediction of releases resulting from HEDD attack requires a method for accounting for the effects of pin pressurization release and the ratio of pin plenum gas release to cask free volume (VR). To account for the effects of VR and to link the two data sources, a simple model has been developed that uses both the SNL data and the GRS data as well as recent test data on aerosols produced in experiments with single pellets subjected to HEDD effects conducted under the aegis of the International Consortium's Working Group on Sabotage of Transport and Storage Casks (WGSTSC). (authors)« less

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Banerjee, Kaushik; Clarity, Justin B; Cumberland, Riley M

This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application.more » Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stafford, Ryan L.; Hinde, Elizabeth; Knight, Mary Jane

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 amore » 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.« less

FFTF disposable solid waste cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomson, J. D.; Goetsch, S. D.

1983-01-01

Disposal of radioactive waste from the Fast Flux Test Facility (FFTF) will utilize a Disposable Solid Waste Cask (DSWC) for the transport and burial of irradiated stainless steel and inconel materials. Retrievability coupled with the desire for minimal facilities and labor costs at the disposal site identified the need for the DSWC. Design requirements for this system were patterned after Type B packages as outlined in 10 CFR 71 with a few exceptions based on site and payload requirements. A summary of the design basis, supporting analytical methods and fabrication practices developed to deploy the DSWC is provided in thismore » paper.« less

Neutron field characterization at the independent spent fuel storage installation of the Trillo nuclear power plant.

PubMed

Campo, Xandra; Méndez, Roberto; Embid, Miguel; Ortego, Alberto; Novo, Manuel; Sanz, Javier

2018-05-01

Neutron fields inside and outside the independent spent fuel storage installation of Trillo Nuclear Power Plant are characterized exhaustively in terms of neutron spectra and ambient dose equivalent, measured by Bonner sphere system and LB6411 monitor. Measurements are consistent with storage casks and building shield characteristics, and also with casks distribution inside the building. Outer values at least five times lower than dose limit for free access area are found. Measurements with LB6411 and spectrometer are consistent with each other. Copyright © 2018 Elsevier Ltd. All rights reserved.

Spent fuel behavior under abnormal thermal transients during dry storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stahl, D.; Landow, M.P.; Burian, R.J.

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less

Mutated CaV2.1 channels dysregulate CASK/P2X3 signaling in mouse trigeminal sensory neurons of R192Q Cacna1a knock-in mice.

PubMed

Gnanasekaran, Aswini; Bele, Tanja; Hullugundi, Swathi; Simonetti, Manuela; Ferrari, Michael D; van den Maagdenberg, Arn M J M; Nistri, Andrea; Fabbretti, Elsa

2013-12-02

ATP-gated P2X3 receptors of sensory ganglion neurons are important transducers of pain as they adapt their expression and function in response to acute and chronic nociceptive signals. The present study investigated the role of calcium/calmodulin-dependent serine protein kinase (CASK) in controlling P2X3 receptor expression and function in trigeminal ganglia from Cacna1a R192Q-mutated knock-in (KI) mice, a genetic model for familial hemiplegic migraine type-1. KI ganglion neurons showed more abundant CASK/P2X3 receptor complex at membrane level, a result that likely originated from gain-of-function effects of R192Q-mutated CaV2.1 channels and downstream enhanced CaMKII activity. The selective CaV2.1 channel blocker ω-Agatoxin IVA and the CaMKII inhibitor KN-93 were sufficient to return CASK/P2X3 co-expression to WT levels. After CASK silencing, P2X3 receptor expression was decreased in both WT and KI ganglia, supporting the role of CASK in P2X3 receptor stabilization. This process was functionally observed as reduced P2X3 receptor currents. We propose that, in trigeminal sensory neurons, the CASK/P2X3 complex has a dynamic nature depending on intracellular calcium and related signaling, that are enhanced in a transgenic mouse model of genetic hemiplegic migraine.

Cosmic ray muons for spent nuclear fuel monitoring

NASA Astrophysics Data System (ADS)

Chatzidakis, Stylianos

There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks from muon measurements. A combination of muon scattering and muon transmission imaging can improve resolution and thus a missing fuel assembly can be identified for vertical and horizontal dry casks. The apparent separation of the images reveals that the muon scattering and transmission can be used for discrimination between casks, satisfying the diversion criteria set by IAEA.

Horizontal modular dry irradiated fuel storage system

DOEpatents

Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

1988-01-01

A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

Improvement of operational safety of dual-purpose transport packaging set for naval SNF in storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guskov, Vladimir; Korotkov, Gennady; Barnes, Ella

2007-07-01

Available in abstract form only. Full text of publication follows: In recent ten years a new technology of management of irradiated nuclear fuel (SNF) at the final stage of fuel cycle has been intensely developing on a basis of a new type of casks used for interim storage of SNF and subsequent transportation therein to the place of processing, further storage or final disposal. This technology stems from the concept of a protective cask which provides preservation of its content (SNF) and fulfillment of all other safety requirements for storage and transportation of SNF. Radiation protection against emissions and non-distributionmore » of activity outside the cask is ensured by physical barriers, i.e. all-metal or composite body, shells, inner cavities for irradiated fuel assemblies (SFA), lids with sealing systems. Residual heat release of SFA is discharged to the environment by natural way: through emission and convection of surrounding air. By now more than 100 dual purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPS). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK- 108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. Within implementation of the international 1.1- 2 project Development of drying technology for the cask TUK-108/1 intended for naval SNF under the Program, there has been developed the technology of preparation of the cask for long-term storage of SNF in TUK-108/1, the design of a mobile TUK-108/1 drying facility; a pilot facility has been manufactured. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in no-108/1, design features of the mobile drying facility, results of tests of the pilot facility at the Far Eastern plant Zvezda. (authors)« less

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

DOE PAGES

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

2016-10-22

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 article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena

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 article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koji Shirai; Jyunichi Tani; Taku Arai

2008-10-01

Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and anglemore » of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.« less

Mutated CaV2.1 channels dysregulate CASK/P2X3 signaling in mouse trigeminal sensory neurons of R192Q Cacna1a knock-in mice

PubMed Central

2013-01-01

Background ATP-gated P2X3 receptors of sensory ganglion neurons are important transducers of pain as they adapt their expression and function in response to acute and chronic nociceptive signals. The present study investigated the role of calcium/calmodulin-dependent serine protein kinase (CASK) in controlling P2X3 receptor expression and function in trigeminal ganglia from Cacna1a R192Q-mutated knock-in (KI) mice, a genetic model for familial hemiplegic migraine type-1. Results KI ganglion neurons showed more abundant CASK/P2X3 receptor complex at membrane level, a result that likely originated from gain-of-function effects of R192Q-mutated CaV2.1 channels and downstream enhanced CaMKII activity. The selective CaV2.1 channel blocker ω-Agatoxin IVA and the CaMKII inhibitor KN-93 were sufficient to return CASK/P2X3 co-expression to WT levels. After CASK silencing, P2X3 receptor expression was decreased in both WT and KI ganglia, supporting the role of CASK in P2X3 receptor stabilization. This process was functionally observed as reduced P2X3 receptor currents. Conclusions We propose that, in trigeminal sensory neurons, the CASK/P2X3 complex has a dynamic nature depending on intracellular calcium and related signaling, that are enhanced in a transgenic mouse model of genetic hemiplegic migraine. PMID:24294842

10 CFR 72.248 - Safety analysis report updating.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety analysis report updating. 72.248 Section 72.248 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF... Approval of Spent Fuel Storage Casks § 72.248 Safety analysis report updating. (a) Each certificate holder...

Final Technical Report: Imaging a Dry Storage Cask with Cosmic Ray Muons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Haori; Hayward, Jason; Chichester, David

The goal of this project is to build a scaled prototype system for monitoring used nuclear fuel (UNF) dry storage casks (DSCs) through cosmic ray muon imaging. Such a system will have the capability of verifying the content inside a DSC without opening it. Because of the growth of the nuclear power industry in the U.S. and the policy decision to ban reprocessing of commercial UNF, the used fuel inventory at commercial reactor sites has been increasing. Currently, UNF needs to be moved to independent spent fuel storage installations (ISFSIs), as its inventory approaches the limit on capacity of on-sitemore » wet storage. Thereafter, the fuel will be placed in shipping containers to be transferred to a final disposal site. The ISFSIs were initially licensed as temporary facilities for ~20-yr periods. Given the cancellation of the Yucca mountain project and no clear path forward, extended dry-cask storage (~100 yr.) at ISFSIs is very likely. From the point of view of nuclear material protection, accountability and control technologies (MPACT) campaign, it is important to ensure that special nuclear material (SNM) in UNF is not stolen or diverted from civilian facilities for other use during the extended storage.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

HOLLENBECK, R.G.

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

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

NASA Astrophysics Data System (ADS)

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.

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ibarra, Luis; Medina, Ricardo; Yang, Haori

This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated withmore » hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.« less

Binding of Y-P30 to Syndecan 2/3 Regulates the Nuclear Localization of CASK

PubMed Central

Landgraf, Peter; Mikhaylova, Marina; Macharadze, Tamar; Borutzki, Corinna; Zenclussen, Ana-Claudia; Wahle, Petra; Kreutz, Michael R.

2014-01-01

The survival promoting peptide Y-P30 has documented neuroprotective effects as well as cell survival and neurite outgrowth promoting activity in vitro and in vivo. Previous work has shown that multimerization of the peptide with pleiotrophin (PTN) and subsequent binding to syndecan (SDC) -2 and -3 is involved in its neuritogenic effects. In this study we show that Y-P30 application regulates the nuclear localization of the SDC binding partner Calcium/calmodulin-dependent serine kinase (CASK) in neuronal primary cultures during development. In early development at day in vitro (DIV) 8 when mainly SDC-3 is expressed supplementation of the culture medium with Y-P30 reduces nuclear CASK levels whereas it has the opposite effect at DIV 18 when SDC-2 is the dominant isoform. In the nucleus CASK regulates gene expression via its association with the T-box transcription factor T-brain-1 (Tbr-1) and we indeed found that gene expression of downstream targets of this complex, like the GluN2B NMDA-receptor, exhibits a corresponding down- or up-regulation at the mRNA level. The differential effect of Y-P30 on the nuclear localization of CASK correlates with its ability to induce shedding of the ectodomain of SDC-2 but not -3. shRNA knockdown of SDC-2 at DIV 18 and SDC-3 at DIV 8 completely abolished the effect of Y-P30 supplementation on nuclear CASK levels. During early development a protein knockdown of SDC-3 also attenuated the effect of Y-P30 on axon outgrowth. Taken together these data suggest that Y-P30 can control the nuclear localization of CASK in a SDC-dependent manner. PMID:24498267

Used fuel rail shock and vibration testing options analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Steven B.; Best, Ralph E.; Klymyshyn, Nicholas A.

2014-09-25

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placedmore » in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.« less

Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

DOE Office of Scientific and Technical Information (OSTI.GOV)

LECHELT, J.A.

2000-10-17

The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System,more » Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix.« less

The scaffold protein calcium/calmodulin-dependent serine protein kinase controls ATP release in sensory ganglia upon P2X3 receptor activation and is part of an ATP keeper complex.

PubMed

Bele, Tanja; Fabbretti, Elsa

2016-08-01

P2X3 receptors, gated by extracellular ATP, are expressed by sensory neurons and are involved in peripheral nociception and pain sensitization. The ability of P2X3 receptors to transduce extracellular stimuli into neuronal signals critically depends on the dynamic molecular partnership with the calcium/calmodulin-dependent serine protein kinase (CASK). The present work used trigeminal sensory neurons to study the impact that activation of P2X3 receptors (evoked by the agonist α,β-meATP) has on the release of endogenous ATP and how CASK modulates this phenomenon. P2X3 receptor function was followed by ATP efflux via Pannexin1 (Panx1) hemichannels, a mechanism that was blocked by the P2X3 receptor antagonist A-317491, and by P2X3 silencing. ATP efflux was enhanced by nerve growth factor, a treatment known to potentiate P2X3 receptor function. Basal ATP efflux was not controlled by CASK, and carbenoxolone or Pannexin silencing reduced ATP release upon P2X3 receptor function. CASK-controlled ATP efflux followed P2X3 receptor activity, but not depolarization-evoked ATP release. Molecular biology experiments showed that CASK was essential for the transactivation of Panx1 upon P2X3 receptor activation. These data suggest that P2X3 receptor function controls a new type of feed-forward purinergic signaling on surrounding cells, with consequences at peripheral and spinal cord level. Thus, P2X3 receptor-mediated ATP efflux may be considered for the future development of pharmacological strategies aimed at containing neuronal sensitization. P2X3 receptors are involved in sensory transduction and associate to CASK. We have studied in primary sensory neurons the molecular mechanisms downstream P2X3 receptor activation, namely ATP release and partnership with CASK or Panx1. Our data suggest that CASK and P2X3 receptors are part of an ATP keeper complex, with important feed-forward consequences at peripheral and central level. © 2016 International Society for Neurochemistry.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-07

...-235, clarify the requirements of reconstituted fuel assemblies, add requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding, delete use of nitrogen for draining...

Comparative changes in color features and pigment composition of red wines aged in oak and cherry wood casks.

PubMed

Chinnici, Fabio; Natali, Nadia; Sonni, Francesca; Bellachioma, Attilio; Riponi, Claudio

2011-06-22

The color features and the evolution of both the monomeric and the derived pigments of red wines aged in oak and cherry 225 L barriques have been investigated during a four months period. For cherry wood, the utilization of 1000 L casks was tested as well. The use of cherry casks resulted in a faster evolution of pigments with a rapid decline of monomeric anthocyanins and a quick augmentation formation of derived and polymeric compounds. At the end of the aging, wines stored in oak and cherry barriques lost, respectively, about 20% and 80% of the initial pigment amount, while in the 1000 L cherry casks, the same compounds diminished by about 60%. Ethyl-bridged adducts and vitisins were the main class of derivatives formed, representing up to 25% of the total pigment amount in the cherry aged samples. Color density augmented in both the oak and cherry wood aged samples, but the latter had the highest values of this parameter. Because of the highly oxidative behavior of the cherry barriques, the use of larger casks (e.g., 1000 L) is proposed in the case of prolonged aging times.

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koji Shirai

2006-04-01

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Radiation Templates of Spent Fuel in Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vanier, Peter

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, butmore » baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.« less

Spent fuel cask handling at an operating nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pal, A.C.

1988-01-01

The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices atmore » all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant.« less

Modification and benchmarking of SKYSHINE-III for use with ISFSI cask arrays

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hertel, N.E.; Napolitano, D.G.

1997-12-01

Dry cask storage arrays are becoming more and more common at nuclear power plants in the United States. Title 10 of the Code of Federal Regulations, Part 72, limits doses at the controlled area boundary of these independent spent-fuel storage installations (ISFSI) to 0.25 mSv (25 mrem)/yr. The minimum controlled area boundaries of such a facility are determined by cask array dose calculations, which include direct radiation and radiation scattered by the atmosphere, also known as skyshine. NAC International (NAC) uses SKYSHINE-III to calculate the gamma-ray and neutron dose rates as a function of distance from ISFSI arrays. In thismore » paper, we present modifications to the SKYSHINE-III that more explicitly model cask arrays. In addition, we have benchmarked the radiation transport methods used in SKYSHINE-III against {sup 60}Co gamma-ray experiments and MCNP neutron calculations.« less

75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-15

... Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415- 6219..., 11555 Rockville Pike, Rockville, Maryland. NRC's Agencywide Documents Access and Management System... M. McCausland, Office of Federal and State Materials and Environmental Management Programs, U.S...

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Fire resistant nuclear fuel cask

DOEpatents

Heckman, Richard C.; Moss, Marvin

1979-01-01

The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technicalmore » basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE PAGES

Vaccaro, S.; Gauld, I. C.; Hu, J.; ...

2018-01-31

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Gauld, I. C.; Hu, J.

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. Finally, the results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.« less

Advancing the Fork detector for quantitative spent nuclear fuel verification

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Gauld, I. C.; Hu, J.; De Baere, P.; Peterson, J.; Schwalbach, P.; Smejkal, A.; Tomanin, A.; Sjöland, A.; Tobin, S.; Wiarda, D.

2018-04-01

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. The results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.

PATRAM '80. Proceedings. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huebner, H.W.

1980-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-09

... definitions for Damaged Fuel Assembly and Transfer Operations; add definitions for Fuel Class and Reconstituted Fuel Assembly; add Combustion Engineering 16x16 class fuel assemblies as authorized contents...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, J.H.; Chipley, K.K.; Nelms, H.A.

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.

Ageing of a neutron shielding used in transport/storage casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nizeyiman, Fidele; Alami, Aatif; Issard, Herve

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.

75 FR 34181 - Connecticut Yankee Atomic Power Company, Haddam Neck Plant, Independent Spent Fuel Storage...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

... Specification (TS) Surveillance Requirement 3.1.6.1 to verify the operability of the concrete cask heat removal... Specification (TS) Surveillance Requirement 3.1.6.1 to verify the operability of the concrete cask heat removal...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-14

... Trussell, Office of Federal and State Materials and Environmental Management Programs, U.S. Nuclear... Management System (ADAMS): Publicly available documents created or received at the NRC are available online... protection of public health and safety continues to be ensured. The direct final rule will become effective...

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

NASA Technical Reports Server (NTRS)

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

1982-01-01

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

Used fuel extended storage security and safeguards by design roadmap

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Lindgren, Eric Richard; Jones, Robert

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. Amore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Winston, Philip Lon

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

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

NASA Astrophysics Data System (ADS)

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

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

Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greiner, Miles

Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding ismore » likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.« less

Proliferation resistance assessment of various methods of spent nuclear fuel storage and disposal

NASA Astrophysics Data System (ADS)

Kollar, Lenka

Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for the spent fuel pool, 0.42 for dry cask storage, 0.36 for the operating geological repository, and 0.28 for the closed geological repository. Therefore, the spent fuel pool is currently the most proliferation resistant method for storing spent fuel. The extrinsic attributes, mainly involving safeguards measures, affect the total PR value the most. As a result, several recommendations are made to improve the proliferation resistance of spent fuel. These recommendations include employing more advanced safeguards measures, such as verification techniques and remote monitoring, for dry cask storage and the geological repository. Dry cask storage facilities should also be located at the plant and in a secure building to minimize the proliferation risk. Finally, the cost-benefit analysis of increased safeguards needs to be considered. Taking these recommendations into account, the PR values of dry cask storage and the closed geological would be significantly increased, to 0.57 and 0.51, respectively. As a result, with increased safeguards to the safeguards level of the spent fuel pool, dry cask storage would be the most proliferation resistant method to store spent fuel. Therefore, the IAEA should continue to develop remote monitoring and cask storage verification techniques in order to improve the proliferation resistance of spent fuel.

10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 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...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 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...

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Development of a novel ultrasonic temperature probe for long-term monitoring of dry cask storage systems

NASA Astrophysics Data System (ADS)

Bakhtiari, S.; Wang, K.; Elmer, T. W.; Koehl, E.; Raptis, A. C.

2013-01-01

With the recent cancellation of the Yucca Mountain repository and the limited availability of wet storage utilities for spent nuclear fuel (SNF), more attention has been directed toward dry cask storage systems (DCSSs) for long-term storage of SNF. Consequently, more stringent guidelines have been issued for the aging management of dry storage facilities that necessitate monitoring of the conditions of DCSSs. Continuous health monitoring of DCSSs based on temperature variations is one viable method for assessing the integrity of the system. In the present work, a novel ultrasonic temperature probe (UTP) is being tested for long-term online temperature monitoring of DCSSs. Its performance was evaluated and compared with type N thermocouple (NTC) and resistance temperature detector (RTD) using a small-scale dry storage canister mockup. Our preliminary results demonstrate that the UTP system developed at Argonne is able to achieve better than 0.8 °C accuracy, tested at temperatures of up to 400 °C. The temperature resolution is limited only by the sampling rate of the current system. The flexibility of the probe allows conforming to complex geometries thus making the sensor particularly suited to measurement scenarios where access is limited.

Benchmarking Data for the Proposed Signature of Used Fuel Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

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 simulationmore » with simulated used nuclear fuel.« less

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Translation and evaluation of the Cultural Awareness Scale for Korean nursing students.

PubMed

Oh, Hyunjin; Lee, Jung-ah; Schepp, Karen G

2015-02-20

To evaluate the effectiveness of a curriculum for achieving high levels of cultural competence, we need to be able to assess education intended to enhance cultural competency skills. We therefore translated the Cultural Awareness Scale (CAS) into Korean (CAS-K). The purpose of this study was to evaluate the cross-cultural applicability and psychometric properties of the CAS-K, specifically its reliability and validity. A cross-sectional descriptive design was used to conduct the evaluation. A convenience sample of 495 nursing students was recruited from four levels of nursing education within four universities in the city of Daejeon, South Korea. This study provided beginning evidence of the validity and reliability of the CAS-K and the cross-cultural applicability of the concepts underlying this instrument. Cronbach's alpha ranged between 0.59 and 0.86 (overall 0.89) in the tests of internal consistency. Cultural competency score prediction of the experience of travel abroad (r=0.084) and the perceived need for cultural education (r=0.223) suggested reasonable criterion validity. Five factors with eigenvalues >1.0 were extracted, accounting for 55.58% of the variance; two retained the same items previously identified for the CAS. The CAS-K demonstrated satisfactory validity and reliability in measuring cultural awareness in this sample of Korean nursing students. The revised CAS-K should be tested for its usability in curriculum evaluation and its applicability as a guide for teaching cultural awareness among groups of Korean nursing students.

Detection of Missing Assemblies and Estimation of the Scattering Densities in a VSC-24 Dry Storage Cask with Cosmic-Ray-Muon-Based Computed Tomography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Zhengzhi; Hayward, Jason; Liao, Can

We report that highly energetic, cosmic-ray muons can penetrate a dry storage cask and yield information about the material inside it by making use of the physics of multiple Coulomb scattering. Work by others has shown this information may be used for verification of dry storage cask contents after continuity of knowledge has been lost. In our modeling and simulation approach, we use ideal planar radiation detectors to record the trajectories and momentum of both incident and exiting cosmic ray muons; this choice allows us to demonstrate the fundamental limit of the technology for a particular measurement and reconstruction method.more » In a method analogous to computed tomography with the attenuation coefficient replaced by scattering density, we apply a filtered back projection algorithm in order to reconstruct the geometry in modeled scenarios for a VSC-24 concrete-walled cask. We also report on our attempt to estimate material-specific information. A scenario where one of the middle four spent nuclear fuel assemblies is missing—undetectable with a simple PoCA-based approach—is expected to be detectable with a CT-based approach. Moreover, a trickier scenario where one or more assemblies is replaced by a dummy assembly is put forward. Lastly, in this case, we expect that this dry storage cask should be found to be not as declared based on our simulation and reconstruction results.« less

Detection of Missing Assemblies and Estimation of the Scattering Densities in a VSC-24 Dry Storage Cask with Cosmic-Ray-Muon-Based Computed Tomography

DOE PAGES

Liu, Zhengzhi; Hayward, Jason; Liao, Can; ...

2017-08-01

We report that highly energetic, cosmic-ray muons can penetrate a dry storage cask and yield information about the material inside it by making use of the physics of multiple Coulomb scattering. Work by others has shown this information may be used for verification of dry storage cask contents after continuity of knowledge has been lost. In our modeling and simulation approach, we use ideal planar radiation detectors to record the trajectories and momentum of both incident and exiting cosmic ray muons; this choice allows us to demonstrate the fundamental limit of the technology for a particular measurement and reconstruction method.more » In a method analogous to computed tomography with the attenuation coefficient replaced by scattering density, we apply a filtered back projection algorithm in order to reconstruct the geometry in modeled scenarios for a VSC-24 concrete-walled cask. We also report on our attempt to estimate material-specific information. A scenario where one of the middle four spent nuclear fuel assemblies is missing—undetectable with a simple PoCA-based approach—is expected to be detectable with a CT-based approach. Moreover, a trickier scenario where one or more assemblies is replaced by a dummy assembly is put forward. Lastly, in this case, we expect that this dry storage cask should be found to be not as declared based on our simulation and reconstruction results.« less

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

NASA Astrophysics Data System (ADS)

Kelley, Ryan P.

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

A&M. Radioactive parts security storage area, heat removal storage casks. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

A&M. Radioactive parts security storage area, heat removal storage casks. Plan, section, and details. Ralph M. Parsons 1480-7 ANP/GE-3-720-S-1. Date: November 1958. Approved by INEEL Classification Office for public release. INEEL index no. 034-0720-60-693-107459 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

... the requirements of reconstituted fuel assemblies; add requirements to qualify metal matrix composite... requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding; clarify the... requirements to qualify metal matrix composite neutron absorbers with integral aluminum cladding; clarify the...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.

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 datamore » 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.« less

10 CFR 72.248 - Safety analysis report updating.

Code of Federal Regulations, 2014 CFR

2014-01-01

... appropriate, the last update to the FSAR under this section. The update shall include the effects 1 of: 1... for a spent fuel storage cask design shall update periodically, as provided in paragraph (b) of this... the issued Certificate of Compliance (CoC). (b) Each update shall contain all the changes necessary to...

10 CFR 72.248 - Safety analysis report updating.

Code of Federal Regulations, 2013 CFR

2013-01-01

... appropriate, the last update to the FSAR under this section. The update shall include the effects 1 of: 1... for a spent fuel storage cask design shall update periodically, as provided in paragraph (b) of this... the issued Certificate of Compliance (CoC). (b) Each update shall contain all the changes necessary to...

10 CFR 72.248 - Safety analysis report updating.

Code of Federal Regulations, 2012 CFR

2012-01-01

... appropriate, the last update to the FSAR under this section. The update shall include the effects 1 of: 1... for a spent fuel storage cask design shall update periodically, as provided in paragraph (b) of this... the issued Certificate of Compliance (CoC). (b) Each update shall contain all the changes necessary to...

10 CFR 72.248 - Safety analysis report updating.

Code of Federal Regulations, 2011 CFR

2011-01-01

... appropriate, the last update to the FSAR under this section. The update shall include the effects 1 of: 1... for a spent fuel storage cask design shall update periodically, as provided in paragraph (b) of this... the issued Certificate of Compliance (CoC). (b) Each update shall contain all the changes necessary to...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-05

... Rule for CoC Format and Content The petitioner states that amending 10 CFR part 72, subpart L, to... conforming changes be made to 10 CFR 72.13. The petitioner argues that ``[n]ew or amended NRC staff positions... 72, subpart L, be amended to remove the requirement that the empty weight be marked on storage casks...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-12

... will be executed will be added when Dominion Virginia Power, who is part of the Electric Power research... Electric Power Research Institute (EPRI) to document what is planned to be accomplished by the CDP. DOE is... Storage Cask Research and Development Project (CDP) AGENCY: Fuel Cycle Technologies, Office of Nuclear...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J. I.

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 withmore » 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.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J. I.

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 withmore » 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.« less

Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.

2007-07-01

This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical densitymore » product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romano, T.

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 validmore » until October 1, 1999. After this date, an update or upgrade to this document is required.« less

An Investigation into the Transportation of Irradiated Uranium/Aluminum Targets from a Foreign Nuclear Reactor to the Chalk River Laboratories Site in Ontario, Canada - 12249

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clough, Malcolm; Jackson, Austin

2012-07-01

This investigation required the selection of a suitable cask and development of a device to hold and transport irradiated targets from a foreign nuclear reactor to the Chalk River Laboratories in Ontario, Canada. The main challenge was to design and validate a target holder to protect the irradiated HEU-Al target pencils during transit. Each of the targets was estimated to have an initial decay heat of 118 W prior to transit. As the targets have little thermal mass the potential for high temperature damage and possibly melting was high. Thus, the primary design objective was to conceive a target holdermore » to dissipate heat from the targets. Other design requirements included securing the targets during transportation and providing a simple means to load and unload the targets while submerged five metres under water. A unique target holder (patent pending) was designed and manufactured together with special purpose experimental apparatus including a representative cask. Aluminum dummy targets were fabricated to accept cartridge heaters, to simulate decay heat. Thermocouples were used to measure the temperature of the test targets and selected areas within the target holder and test cask. After obtaining test results, calculations were performed to compensate for differences between experimental and real life conditions. Taking compensation into consideration the maximum target temperature reached was 231 deg. C which was below the designated maximum of 250 deg. C. The design of the aluminum target holder also allowed generous clearance to insert and unload the targets. This clearance was designed to close up as the target holder is placed into the cavity of the transport cask. Springs served to retain and restrain the targets from movement during transportation as well as to facilitate conductive heat transfer. The target holder met the design requirements and as such provided data supporting the feasibility of transporting targets over a relatively long period of time. A suitable transport cask was selected and a device for housing irradiated targets for loading, unloading and transportation has been designed, built and validated. The device was successful in meeting all design requirements for this feasibility study. Experiments were conducted with a custom test facility to confirm that the design met the maximum temperature requirements during shipping. Results from tests showed that the peak temperature in the apparatus was 300 deg. C. By compensating for experimental considerations, such as reduced thermal conductivity of the test cask versus that of the actual cask the expected maximum target temperature reduces to 231 deg. C. This is below the designated peak value of 250 deg. C. It can therefore be concluded, based on the content of this paper and from a heat-removal standpoint, the feasibility of transporting targets from a foreign nuclear reactor to Canada is possible, although further testing with irradiated targets and a full size cask would be a recommended next step. (authors)« less

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Bowman, Stephen M; Gauld, Ian C

2015-01-01

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, andmore » it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades are inserted in various locations and at varying degrees during BWR operation based on the reload design. The presence of control blades during depletion hardens the neutron spectrum locally due to both moderator displacement and introduction of a thermal neutron absorber. The reactivity impact of control blade presence is investigated herein, as well as the effect of multiple (continuous and intermittent) exposure periods. The coupled effects of control blade presence on power density, void profile, or burnup profile have not been considered to date but will be addressed in future work.« less

Accident analysis and control options in support of the sludge water system safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

HEY, B.E.

A hazards analysis was initiated for the SWS in July 2001 (SNF-8626, K Basin Sludge and Water System Preliminary Hazard Analysis) and updated in December 2001 (SNF-10020 Rev. 0, Hazard Evaluation for KE Sludge and Water System - Project A16) based on conceptual design information for the Sludge Retrieval System (SRS) and 60% design information for the cask and container. SNF-10020 was again revised in September 2002 to incorporate new hazards identified from final design information and from a What-if/Checklist evaluation of operational steps. The process hazards, controls, and qualitative consequence and frequency estimates taken from these efforts have beenmore » incorporated into Revision 5 of HNF-3960, K Basins Hazards Analysis. The hazards identification process documented in the above referenced reports utilized standard industrial safety techniques (AIChE 1992, Guidelines for Hazard Evaluation Procedures) to systematically guide several interdisciplinary teams through the system using a pre-established set of process parameters (e.g., flow, temperature, pressure) and guide words (e.g., high, low, more, less). The teams generally included representation from the U.S. Department of Energy (DOE), K Basins Nuclear Safety, T Plant Nuclear Safety, K Basin Industrial Safety, fire protection, project engineering, operations, and facility engineering.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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).

The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, C.; Thomas, I.

2013-07-01

This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility,more » the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.« less

National Policy Implications of Storing Nuclear Waste in the Pacific Region,

DTIC Science & Technology

1981-01-01

US Congress, Senate, Committee on Energy and Natural Resources, Pacific Spent Nuclear Fuel Storage , Hearing...selected. 17 One type of shipping cask which has been used to transport spent fuel assemblies to the Nevada Test Site is a leakproof steel cask that can...discussion the following conclusions on the nuclear waste storage issue appear valid. The Reagan decision to reprocess spent fuel has not changed US

Performance of bolted closure joint elastomers under cask aging conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Verst, C.; Sindelar, R.; Skidmore, E.

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 temperaturemore » 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.« less

Dynamic Impact Analyses and Tests of Concrete Overpacks - 13638

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Sanghoon; Cho, Sang-Soon; Kim, Ki-Young

Concrete cask is an option for spent nuclear fuel interim storage which is prevailingly used in US. A concrete cask usually consists of metallic canister which confines the spent nuclear fuel and concrete overpack. When the overpack undergoes a severe missile impact which might be caused by a tornado or an aircraft crash, it should sustain acceptable level of structural integrity so that its radiation shielding capability and the retrievability of canister are maintained. Missile impact against a concrete overpack involves two damage modes, local damage and global damage. Local damage of concrete is usually evaluated by empirical formulas whilemore » the global damage is evaluated by finite element analysis. In many cases, those two damage modes are evaluated separately. In this research, a series of numerical simulations are performed using finite element analysis to evaluate the global damage of concrete overpack as well as its local damage under high speed missile impact. We consider two types of concrete overpack, one with steel in-cased concrete without reinforcement and the other with partially-confined reinforced concrete. The numerical simulation results are compared with test results and it is shown that appropriate modeling of material failure is crucial in this analysis and the results are highly dependent on the choice of failure parameters. (authors)« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-06

... of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule... technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power... NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI75 [NRC-2009-0538] List of Approved Spent...

75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-15

... of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule... technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power... NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 [NRC-2010-0140] RIN 3150-AI86 List of Approved Spent...

77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-27

... #0; #0;Rules and Regulations #0; Federal Register #0; #0; #0;This section of the FEDERAL REGISTER contains regulatory documents #0;having general applicability and legal effect, most of which are keyed #0;to and codified in the Code of Federal Regulations, which is published #0;under 50 titles pursuant to...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-25

...;Prices of new books are listed in the first FEDERAL REGISTER issue of each #0;week. #0; #0; #0; #0;#0... ADAMS Search.'' For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference... not have a significant economic impact on a substantial number of small entities. This final rule...

76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-14

... various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water Reactor baskets... add various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water Reactor....1.1 to add various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Luna, Robert Earl

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 thatmore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingham, J.G.

The IDENT 1578 container, which is a 110-in. long 5.5-in. OD tube, is designed for shipping FFTF fuel elements in T-3 casks between HEDL, HFEF, and other laboratories. The thermal analysis was conducted to evaluate whether or not the container satisfies its thermal design criteria (handle a decay heat load of 600 watts, max fuel pin cladding temperature not exceeding 800/sup 0/F).

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Fuel shipment experience, fuel movements from the BMI-1 transport cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

Congeners in sugar cane spirits aged in casks of different woods.

PubMed

Bortoletto, Aline M; Alcarde, André R

2013-08-15

The profile of volatile compounds and aging markers in sugar cane spirits aged for 36 months in casks made of 10 types of wood were studied. The ethanol content, volatile acidity, aldehydes, esters, higher alcohols, and methanol were determined. In addition, gallic, vanilic and syringic acids, siringaldehyde, coniferaldehyde, sinapaldehyde, vanillin, 5-hydroxymethylfurfural and furfural were identified and quantified. The profile of volatile compounds characterised aging in each type of wood. The beverage aged in oak cask achieved the highest contents of maturation-related congeners. The Brazilian woods, similar to oak, were jequitibá rosa and cerejeira, which presented the highest contents of some maturation-related compounds, such as vanillin, vanilic acid, syringaldehyde and sinapaldehyde. Although oak wood conferred more chemical complexity to the beverage, Brazilian woods, singly or complementarily, present potential for spirit characterisation and for improving the quality of sugar cane spirits. Copyright © 2013 Elsevier Ltd. All rights reserved.

Characterization of Hydrogen Embrittled Zircaloy-4 by Using a Van de Graaff Particle Accelerator

NASA Astrophysics Data System (ADS)

Budd, John

2013-04-01

On-site, dry cask storage was originally by the intended to be a short-term solution for holding spent nuclear fuel. Due to the lack of a permanent storage facility, the nuclear power industry seeks to assess the effective lifetime of the casks. One issue which could compromise cask integrity is Hydrogen embrittlement. This phenomenon occurs in the Zircaloy-4 fuel-rod cladding and is caused by the formation of Zirconium hydrides. Over time, thermal stresses caused by the heat from reactions of the stored nuclear fuel could result in significant breaches of the cladding. Our group at Texas A&M University- Kingsville is conducting experiments to aid in determining when such breaches will occur. We will irradiate samples of the alloy with protons of energies up to 400 keV using a Van de Graaff particle accelerator. Once irradiated, their properties will be characterized using scanning electron microscopy and Vickers hardness tests.

Radionuclide production and dose rate estimation during the commissioning of the W-Ta spallation target

NASA Astrophysics Data System (ADS)

Yu, Q. Z.; Liang, T. J.

2018-06-01

China Spallation Neutron Source (CSNS) is intended to begin operation in 2018. CSNS is an accelerator-base multidisciplinary user facility. The pulsed neutrons are produced by a 1.6GeV short-pulsed proton beam impinging on a W-Ta spallation target, at a beam power of100 kW and a repetition rate of 25 Hz. 20 neutron beam lines are extracted for the neutron scattering and neutron irradiation research. During the commissioning and maintenance scenarios, the gamma rays induced from the W-Ta target can cause the dose threat to the personal and the environment. In this paper, the gamma dose rate distributions for the W-Ta spallation are calculated, based on the engineering model of the target-moderator-reflector system. The shipping cask is analyzed to satisfy the dose rate limit that less than 2 mSv/h at the surface of the shipping cask. All calculations are performed by the Monte carlo code MCNPX2.5 and the activation code CINDER’90.

Develop an piezoelectric sensing based on SHM system for nuclear dry storage system

NASA Astrophysics Data System (ADS)

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.

75 FR 33853 - Maine Yankee Atomic Power Company; Independent Spent Fuel Storage Installation; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-15

...) Surveillance Requirement 3.1.6.1 to verify the operability of the concrete cask heat removal system to maintain... Amendment No. 5 for one storage canister at the MY ISFSI. The affected storage canister had a heat load of 9..., and the LCO 3.1.4 time limit for a canister [[Page 33855

Management of the Cs/Sr Capsule Project at the Hanford Site. Technology Readiness Assessment Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The Federal Project Director (FPD) for the U.S. Department of Energy (DOE), Richland Operations Office (RL) Waste Management and D&D Division (WMD) requested a Technology Readiness Assessment (TRA) for the Management of the Cesium/Strontium Capsule Storage Project (MCSCP) at the Waste Encapsulation and Storage Facility (WESF) on the Hanford Site in Washington State. The MCSCP CD-1 TRA was performed by a team selected in collaboration between the Office of Environmental Management (EM) Chief Engineer (EM-3.3) and RL, WMD FPD. The TRA Team included subject matter and technical experts having experience in cask storage, process engineering, and system design who weremore » independent of the MCSCP, and the team was led by the Director of Operations and Processes from the EM Chief Engineer's Office (EM-3.32). Movement of the Cs/Sr capsules to dry storage, based on information from the conceptual design, involves (1) capsule packaging, (2) capsule transfer, and (3) capsule storage. The project has developed a conceptual process, described in 30059-R-02, "NAC Conceptual Design Report for the Management of the Cesium and Strontium Capsules Project", which identifies the five major activities in the process to complete the transfer from storage pool to pad-mounted cask storage. The process, shown schematically in Figure 1, is comprised of the following process steps: (1) loading capsules into the UCS; (2) UCS processing; (3) UCS insertion into the TSC Basket; (4) cask transport from WESF to CSA and (5) extended storage at the CSA.« less

Evaluation of microwave cavity gas sensor for in-vessel monitoring of dry cask storage systems

NASA Astrophysics Data System (ADS)

Bakhtiari, S.; Gonnot, T.; Elmer, T.; Chien, H.-T.; Engel, D.; Koehl, E.; Heifetz, A.

2018-04-01

Results are reported of research activities conducted at Argonne to assess the viability of microwave resonant cavities for extended in-vessel monitoring of dry cask storage system (DCSS) environment. One of the gases of concern to long-term storage in canisters is water vapor, which appears due to evaporation of residual moisture from incompletely dried fuel assembly. Excess moisture could contribute to corrosion and deterioration of components inside the canister, which would in turn compromise maintenance and safe transportation of such systems. Selection of the sensor type in this work was based on a number of factors, including good sensitivity, fast response time, small form factor and ruggedness of the probing element. A critical design constraint was the capability to mount and operate the sensor using the existing canister penetrations-use of existing ports for thermocouple lances. Microwave resonant cavities operating at select resonant frequency matched to the rotational absorption line of the molecule of interest offer the possibility of highly sensitive detection. In this study, two prototype K-band microwave cylindrical cavities operating at TE01n resonant modes around the 22 GHz water absorption line were developed and tested. The sensors employ a single port for excitation and detection and a novel dual-loop inductive coupling for optimized excitation of the resonant modes. Measurement of the loaded and unloaded cavity quality factor was obtained from the S11 parameter. The acquisition and real-time analysis of data was implemented using software based tools developed for this purpose. The results indicate that the microwave humidity sensors developed in this work could be adapted to in-vessel monitoring applications that require few parts-per-million level of sensitivity. The microwave sensing method for detection of water vapor can potentially be extended to detection of radioactive fission gases leaking into the interior of the canister through cracks in fuel cladding.

Quantity and management of spent fuel from prototype and research reactors in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

Nuclear cask testing films misleading and misused

DOE Office of Scientific and Technical Information (OSTI.GOV)

Audin, L.

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 asmore » 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.« less

Nuclear cask testing films misleading and misused

DOE Office of Scientific and Technical Information (OSTI.GOV)

Audin, L.

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 asmore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, Eric Benton

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

SLUDGE TREATMENT PROJECT COST COMPARISON BETWEEN HYDRAULIC LOADING AND SMALL CANISTER LOADING CONCEPTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

GEUTHER J; CONRAD EA; RHOADARMER D

2009-08-24

The Sludge Treatment Project (STP) is considering two different concepts for the retrieval, loading, transport and interim storage of the K Basin sludge. The two design concepts under consideration are: (1) Hydraulic Loading Concept - In the hydraulic loading concept, the sludge is retrieved from the Engineered Containers directly into the Sludge Transport and Storage Container (STSC) while located in the STS cask in the modified KW Basin Annex. The sludge is loaded via a series of transfer, settle, decant, and filtration return steps until the STSC sludge transportation limits are met. The STSC is then transported to T Plantmore » and placed in storage arrays in the T Plant canyon cells for interim storage. (2) Small Canister Concept - In the small canister concept, the sludge is transferred from the Engineered Containers (ECs) into a settling vessel. After settling and decanting, the sludge is loaded underwater into small canisters. The small canisters are then transferred to the existing Fuel Transport System (FTS) where they are loaded underwater into the FTS Shielded Transfer Cask (STC). The STC is raised from the basin and placed into the Cask Transfer Overpack (CTO), loaded onto the trailer in the KW Basin Annex for transport to T Plant. At T Plant, the CTO is removed from the transport trailer and placed on the canyon deck. The CTO and STC are opened and the small canisters are removed using the canyon crane and placed into an STSC. The STSC is closed, and placed in storage arrays in the T Plant canyon cells for interim storage. The purpose of the cost estimate is to provide a comparison of the two concepts described.« less

Numerical Estimation of the Spent Fuel Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason

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

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

NASA Astrophysics Data System (ADS)

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.

KSC-2011-6659

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the RTG storage facility at NASA's Kennedy Space Center in Florida, the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission is lowered to the floor of the high bay in preparation for lifting the cask from around the MMRTG. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

Conceptual design of fast-ignition laser fusion reactor FALCON-D

NASA Astrophysics Data System (ADS)

Goto, T.; Someya, Y.; Ogawa, Y.; Hiwatari, R.; Asaoka, Y.; Okano, K.; Sunahara, A.; Johzaki, T.

2009-07-01

A new conceptual design of the laser fusion power plant FALCON-D (Fast-ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast-ignition method can achieve sufficient fusion gain for a commercial operation (~100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5-6 m radius). 1D/2D simulations by hydrodynamic codes showed a possibility of achieving sufficient gain with a laser energy of 400 kJ, i.e. a 40 MJ target yield. The design feasibility of the compact dry wall chamber and the solid breeder blanket system was shown through thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. Moderate electric output (~400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall reactor concept not only reduces several difficulties associated with a liquid wall system but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance period. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R&D issues required for this design are also discussed.

KSC-2011-6646

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- The multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory (MSL) mission, enclosed in a shipping cask in the MMRTG trailer, arrives at the RTG storage facility at NASA's Kennedy Space Center in Florida. During transport, coolant flows through hoses connected to the cask to dissipate any excess heat generated by the MMRTG. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

Synchronous fluorescence spectroscopy for analysis of wine and wine distillates

NASA Astrophysics Data System (ADS)

Andreeva, Ya.; Borisova, E.; Genova, Ts.; Zhelyazkova, Al.; Avramov, L.

2015-01-01

Wine and brandies are multicomponent systems and conventional fluorescence techniques, relying on recording of single emission or excitation spectra, are often insufficient. In such cases synchronous fluorescence spectra can be used for revealing the potential of the fluorescence techniques. The technique is based on simultaneously scanning of the excitation and emission wavelength with constant difference (Δλ) maintained between them. In this study the measurements were made using FluoroLog3 spectrofluorimeter (HORIBA Jobin Yvon, France) and collected for excitation and emission in the wavelength region 220 - 700 nm using wavelength interval Δλ from 10 to 100 nm in 10 nm steps. This research includes the results obtained for brandy and red wine samples. Fluorescence analysis takes advantage in the presence of natural fluorophores in wines and brandies, such as gallic, vanillic, p-coumaric, syringic, ferulic acid, umbelliferone, scopoletin and etc. Applying of synchronous fluorescence spectroscopy for analysis of these types of alcohols allows us to estimate the quality of wines and also to detect adulteration of brandies like adding of a caramel to wine distillates for imitating the quality of the original product aged in oak casks.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsoukalas, Lefteri H.

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.

Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Svitak, F.; Broz, V.; Hrehor, M.

2008-07-15

The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF formore » transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)« less

Recent developments - US spent fuel disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

One of a US utility's major risk factors in continuing to operate a nuclear plant is managing discharged spent fuel. The US Department of Energy (DOE) signed contracts with utilities guaranteeing government acceptance of spent fuel by 1988. However, on December 17, 1992, DOE Secretary Watkins wrote to Sen. J. Bennett Johnston (D-LA), Chairman of the Senate Energy Committee, indicating a reassessment of DOE's programs, the results of which will be presented to Congress in January 1993. He indicated the Department may not be able to meet the 1988 date, because of difficulty in finding a site for the Monitoredmore » Retrievable Storage facility. Watkins indicated that DOE has investigated an interim solution and decided to expedite a program to certify a multi-purpose standardized cask system for spent fuel receipt, storage, transport, and disposal. To meet the expectations of US utilities, DOE is considering a plan to use federal sites for interim storage of the casks. Secretary Watkins recommended the waste program be taken off-budget and put in a revolving fund established to ensure that money already collected from utilities will be available to meet the schedule for completion of the repository.« less

Electromagnetic Acoustic Transducers for Robotic Nondestructive Inspection in Harsh Environments.

PubMed

Choi, Sungho; Cho, Hwanjeong; Lindsey, Matthew S; Lissenden, Cliff J

2018-01-11

Elevated temperature, gamma radiation, and geometric constraints inside dry storage casks for spent nuclear fuel represent a harsh environment for nondestructive inspection of the cask and require that the inspection be conducted with a robotic system. Electromagnetic acoustic transducers (EMATs) using non-contact ultrasonic transduction based on the Lorentz force to excite/receive ultrasonic waves are suited for use in the robotic inspection. Periodic permanent magnet EMATs that actuate/receive shear horizontal guided waves are developed for application to robotic nondestructive inspection of stress corrosion cracks in the heat affected zone of welds in stainless steel dry storage canisters. The EMAT's components are carefully selected in consideration of the inspection environment, and tested under elevated temperature and gamma radiation doses up to 177 °C and 5920 krad, respectively, to evaluate the performance of the EMATs under realistic environmental conditions. The effect of gamma radiation is minimal, but the EMAT's performance is affected by temperatures above 121 °C due to the low Curie temperature of the magnets. Different magnets are needed to operate at 177 °C. The EMAT's capability to detect notches is also evaluated from B-scan measurements on 304 stainless steel welded plate containing surface-breaking notches.

Plasma Membrane Ca2+-ATPase 4 in Murine Epididymis: Secretion of Splice Variants in the Luminal Fluid and a Role in Sperm Maturation1

PubMed Central

Patel, Ramkrishna; Al-Dossary, Amal A.; Stabley, Deborah L.; Barone, Carol; Galileo, Deni S.; Strehler, Emanuel E.; Martin-DeLeon, Patricia A.

2013-01-01

ABSTRACT Plasma membrane Ca2+-ATPase isoform 4 (PMCA4) is the primary Ca2+ efflux pump in murine sperm, where it regulates motility. In Pmca4 null sperm, motility loss results in infertility. We have shown that murine sperm PMCA4b interacts with Ca2+/CaM-dependent serine kinase (CASK) in regulating Ca2+ homeostasis and motility. However, recent work indicated that the bovine PMCA4a splice variant (missing in testis) is epididymally expressed, along with 4b, and may be transferred to sperm. Here we show, via conventional and in situ RT-PCR, that both the splice variants of Pmca4 mRNA are expressed in murine testis and throughout the epididymis. Immunofluorescence localized PMCA4a to the apical membrane of the epididymal epithelium, and Western analysis not only confirmed its presence but showed for the first time that PMCA4a and PMCA4b are secreted in the epididymal luminal fluid (ELF), from which epididymosomes containing PMCA4a were isolated. Flow cytometry indicated the presence of PMCA4a on mature caudal sperm where it was increased ∼5-fold compared to caput sperm (detected by Western blotting) and ∼2-fold after incubation in ELF, revealing in vitro uptake and implicating PMCA4a in epididymal sperm maturation. Coimmunoprecipitation using pan-PMCA4 antibodies, revealed that both variants associate with CASK, suggesting their presence in a complex. Because they have different kinetic properties for Ca2+ transport and different abilities to bind to CASK, our study suggests a mechanism for combining the functional attributes of both PMCA4 variants, leading to heightened efficiency of the pump in the maintenance of Ca2+ homeostasis, which is crucial for normal motility and male fertility. PMID:23699388

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Lumin; Wierschke, Jonathan Brett

2015-04-08

The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised ofmore » boron trioxide and sassolite (H 3BO 3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.« less

Plasma membrane Ca2+-ATPase 4 in murine epididymis: secretion of splice variants in the luminal fluid and a role in sperm maturation.

PubMed

Patel, Ramkrishna; Al-Dossary, Amal A; Stabley, Deborah L; Barone, Carol; Galileo, Deni S; Strehler, Emanuel E; Martin-DeLeon, Patricia A

2013-07-01

Plasma membrane Ca(2+)-ATPase isoform 4 (PMCA4) is the primary Ca(2+) efflux pump in murine sperm, where it regulates motility. In Pmca4 null sperm, motility loss results in infertility. We have shown that murine sperm PMCA4b interacts with Ca(2+)/CaM-dependent serine kinase (CASK) in regulating Ca(2+) homeostasis and motility. However, recent work indicated that the bovine PMCA4a splice variant (missing in testis) is epididymally expressed, along with 4b, and may be transferred to sperm. Here we show, via conventional and in situ RT-PCR, that both the splice variants of Pmca4 mRNA are expressed in murine testis and throughout the epididymis. Immunofluorescence localized PMCA4a to the apical membrane of the epididymal epithelium, and Western analysis not only confirmed its presence but showed for the first time that PMCA4a and PMCA4b are secreted in the epididymal luminal fluid (ELF), from which epididymosomes containing PMCA4a were isolated. Flow cytometry indicated the presence of PMCA4a on mature caudal sperm where it was increased ~5-fold compared to caput sperm (detected by Western blotting) and ~2-fold after incubation in ELF, revealing in vitro uptake and implicating PMCA4a in epididymal sperm maturation. Coimmunoprecipitation using pan-PMCA4 antibodies, revealed that both variants associate with CASK, suggesting their presence in a complex. Because they have different kinetic properties for Ca(2+) transport and different abilities to bind to CASK, our study suggests a mechanism for combining the functional attributes of both PMCA4 variants, leading to heightened efficiency of the pump in the maintenance of Ca(2+) homeostasis, which is crucial for normal motility and male fertility.

Depleted uranium hexafluoride: The source material for advanced shielding systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quapp, W.J.; Lessing, P.A.; Cooley, C.R.

1997-02-01

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

Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

1989-08-01

The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silviera, D.J.; Aaberg, R.L.; Cushing, C.E.

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

Electromagnetic Acoustic Transducers for Robotic Nondestructive Inspection in Harsh Environments

PubMed Central

Choi, Sungho; Cho, Hwanjeong; Lindsey, Matthew S.; Lissenden, Cliff J.

2018-01-01

Elevated temperature, gamma radiation, and geometric constraints inside dry storage casks for spent nuclear fuel represent a harsh environment for nondestructive inspection of the cask and require that the inspection be conducted with a robotic system. Electromagnetic acoustic transducers (EMATs) using non-contact ultrasonic transduction based on the Lorentz force to excite/receive ultrasonic waves are suited for use in the robotic inspection. Periodic permanent magnet EMATs that actuate/receive shear horizontal guided waves are developed for application to robotic nondestructive inspection of stress corrosion cracks in the heat affected zone of welds in stainless steel dry storage canisters. The EMAT’s components are carefully selected in consideration of the inspection environment, and tested under elevated temperature and gamma radiation doses up to 177 °C and 5920 krad, respectively, to evaluate the performance of the EMATs under realistic environmental conditions. The effect of gamma radiation is minimal, but the EMAT’s performance is affected by temperatures above 121 °C due to the low Curie temperature of the magnets. Different magnets are needed to operate at 177 °C. The EMAT’s capability to detect notches is also evaluated from B-scan measurements on 304 stainless steel welded plate containing surface-breaking notches. PMID:29324721

Validation Test Report For The CRWMS Analysis and Logistics Visually Interactive Model Calvin Version 3.0, 10074-Vtr-3.0-00

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Gillespie

2000-07-27

This report describes the tests performed to validate the CRWMS ''Analysis and Logistics Visually Interactive'' Model (CALVIN) Version 3.0 (V3.0) computer code (STN: 10074-3.0-00). To validate the code, a series of test cases was developed in the CALVIN V3.0 Validation Test Plan (CRWMS M&O 1999a) that exercises the principal calculation models and options of CALVIN V3.0. Twenty-five test cases were developed: 18 logistics test cases and 7 cost test cases. These cases test the features of CALVIN in a sequential manner, so that the validation of each test case is used to demonstrate the accuracy of the input to subsequentmore » calculations. Where necessary, the test cases utilize reduced-size data tables to make the hand calculations used to verify the results more tractable, while still adequately testing the code's capabilities. Acceptance criteria, were established for the logistics and cost test cases in the Validation Test Plan (CRWMS M&O 1999a). The Logistics test cases were developed to test the following CALVIN calculation models: Spent nuclear fuel (SNF) and reactivity calculations; Options for altering reactor life; Adjustment of commercial SNF (CSNF) acceptance rates for fiscal year calculations and mid-year acceptance start; Fuel selection, transportation cask loading, and shipping to the Monitored Geologic Repository (MGR); Transportation cask shipping to and storage at an Interim Storage Facility (ISF); Reactor pool allocation options; and Disposal options at the MGR. Two types of cost test cases were developed: cases to validate the detailed transportation costs, and cases to validate the costs associated with the Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) and Regional Servicing Contractors (RSCs). For each test case, values calculated using Microsoft Excel 97 worksheets were compared to CALVIN V3.0 scenarios with the same input data and assumptions. All of the test case results compare with the CALVIN V3.0 results within the bounds of the acceptance criteria. Therefore, it is concluded that the CALVIN V3.0 calculation models and options tested in this report are validated.« less

TREAT neutron-radiography facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, L.J.

1981-01-01

The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This volume contains the interim change notice for sample preparation methods. Covered are: acid digestion for metals analysis, fusion of Hanford tank waste solids, water leach of sludges/soils/other solids, extraction procedure toxicity (simulate leach in landfill), sample preparation for gamma spectroscopy, acid digestion for radiochemical analysis, leach preparation of solids for free cyanide analysis, aqueous leach of solids for anion analysis, microwave digestion of glasses and slurries for ICP/MS, toxicity characteristic leaching extraction for inorganics, leach/dissolution of activated metal for radiochemical analysis, extraction of single-shell tank (SST) samples for semi-VOC analysis, preparation and cleanup of hydrocarbon- containing samples for VOCmore » and semi-VOC analysis, receiving of waste tank samples in onsite transfer cask, receipt and inspection of SST samples, receipt and extrusion of core samples at 325A shielded facility, cleaning and shipping of waste tank samplers, homogenization of solutions/slurries/sludges, and test sample preparation for bioassay quality control program.« less

Pulse Shape Analysis and Discrimination for Silicon-Photomultipliers in Helium-4 Gas Scintillation Neutron Detector

NASA Astrophysics Data System (ADS)

Barker, Cathleen; Zhu, Ting; Rolison, Lucas; Kiff, Scott; Jordan, Kelly; Enqvist, Andreas

2018-01-01

Using natural helium (helium-4), the Arktis 180-bar pressurized gas scintillator is capable of detecting and distinguishing fast neutrons and gammas. The detector has a unique design of three optically separated segments in which 12 silicon-photomultiplier (SiPM) pairs are positioned equilaterally across the detector to allow for them to be fully immersed in the helium-4 gas volume; consequently, no additional optical interfaces are necessary. The SiPM signals were amplified, shaped, and readout by an analog board; a 250 MHz, 14-bit digitizer was used to examine the output pulses from each SiPMpair channel. The SiPM over-voltage had to be adjusted in order to reduce pulse clipping and negative overshoot, which was observed for events with high scintillation production. Pulse shaped discrimination (PSD) was conducted by evaluating three different parameters: time over threshold (TOT), pulse amplitude, and pulse integral. In order to differentiate high and low energy events, a 30ns gate window was implemented to group pulses from two SiPM channels or more for the calculation of TOT. It was demonstrated that pulses from a single SiPM channel within the 30ns window corresponded to low-energy gamma events while groups of pulses from two-channels or more were most likely neutron events. Due to gamma pulses having lower pulse amplitude, the percentage of measured gamma also depends on the threshold value in TOT calculations. Similarly, the threshold values were varied for the optimal PSD methods of using pulse amplitude and pulse area parameters. Helium-4 detectors equipped with SiPMs are excellent for in-the-field radiation measurement of nuclear spent fuel casks. With optimized PSD methods, the goal of developing a fuel cask content monitoring and inspection system based on these helium-4 detectors will be achieved.

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, B. D.

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and themore » mechanical properties of the rods will be tested and analyzed.« less

Piezoelectric wafer active sensors under gamma radiation exposure toward applications for structural health monitoring of nuclear dry cask storage systems

NASA Astrophysics Data System (ADS)

Faisal Haider, Mohammad; Mei, Hanfei; Lin, Bin; Yu, Lingyu; Giurgiutiu, Victor; Lam, Poh-Sang; Verst, Christopher

2018-03-01

Structural health monitoring (SHM) is in urgent need and must be integrated into the nuclear-spent fuel storage systems to guarantee the safe operation. The dry cask storage system (DCSS) is such storage facility, which is licensed for temporary storage for nuclear-spent fuel at the independent spent fuel storage installations (ISFSIs) for certain predetermined period of time. Gamma radiation is one of the major radiation sources near DCSS. Therefore, a detailed experimental investigation was completed on the gamma radiation endurance of piezoelectric wafer active sensors (PWAS) transducers for SHM applications to the DCSS system. The irradiation test was done in a Co-60 gamma irradiator. Lead Zirconate Titanate (PZT) and Gallium Orthophosphate (GaPO4) PWAS transducers were exposed to 40.7 kGy gamma radiation. Total radiation dose was achieved in two different radiation dose rates: (a) slower radiation rate at 0.1 kGy/hr for 20 hours (b) accelerated radiation rate at 1.233 kGy/hr for 32 hours. The total cumulative radiation dose of 40.7 kGy is equivalent to 45 years of operation in DCSS system. Electro-mechanical impedance and admittance (EMIA) signatures and electrical capacitance were measured to evaluate the PWAS performance after each gamma radiation exposure. The change in resonance frequency of PZT-PWAS transducer for both in-plane and thickness mode was observed. The GaPO4-PWAS EMIA spectra do not show a significant shift in resonance frequency after gamma irradiation exposure. Radiation endurance of new high-temperature HPZ-HiT PWAS transducer was also evaluated. The HPZ-HiT transducers were exposed to gamma radiation at 1.233 kGy/hr for 160 hours with 80 hours interval. Therefore, the total accumulated gamma radiation dose is 184 kGy. No significant change in impedance spectra was observed due to gamma radiation exposure.

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

2015-01-01

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members acceptmore » the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).« less

Draft Geologic Disposal Requirements Basis for STAD Specification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilgen, Anastasia G.; Bryan, Charles R.; Hardin, Ernest

2015-03-25

This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussedmore » here.« less

Experiences with welding multi-assembly sealed baskets at Palisades

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agace, S.; Worrell, S.; Stewart, L.

1995-12-01

Four utilities were using operational canister-based dry storage facilities at year-end, and seven more have contracts to establish similar facilities. Consumers Power`s Palisades Nuclear Power Plant has successfully completed loading its eighth dry storage canister with the Ventilated Storage Cask (VSC) system, under license to Sierra Nuclear Corporation. The VSC has a Multi-Assembly Sealed Basket (MSB) containing 24 specially-selected and aged spent fuel assemblies. MSB closure occurs when two independent lids are welded at the utility. The canister wall and lids are SA-516 Grade 70 carbon steel. This paper discusses the welding system design, closure operations and MSB closure operationsmore » at Palisades.« less

27 CFR 26.206 - Marking packages and cases.

Code of Federal Regulations, 2010 CFR

2010-04-01

..., rectifier, or bottler shall serially number each case, barrel, cask, or similar container of distilled... distiller, rectifier, or bottler shall plainly print, stamp, or stencil with durable coloring material, in...

RH-TRU Waste Shipments from Battelle Columbus Laboratories to the Hanford Nuclear Facility for Interim Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eide, J.; Baillieul, T. A.; Biedscheid, J.

2003-02-26

Battelle Columbus Laboratories (BCL), located in Columbus, Ohio, must complete decontamination and decommissioning (D&D) activities for nuclear research buildings and grounds by 2006, as directed by Congress. Most of the resulting waste (approximately 27 cubic meters [m3]) is remote-handled (RH) transuranic (TRU) waste destined for disposal at the Waste Isolation Pilot Plant (WIPP). The BCL, under a contract to the U.S. Department of Energy (DOE) Ohio Field Office, has initiated a plan to ship the TRU waste to the DOE Hanford Nuclear Facility (Hanford) for interim storage pending the authorization of WIPP for the permanent disposal of RH-TRU waste. Themore » first of the BCL RH-TRU waste shipments was successfully completed on December 18, 2002. This BCL shipment of one fully loaded 10-160B Cask was the first shipment of RH-TRU waste in several years. Its successful completion required a complex effort entailing coordination between different contractors and federal agencies to establish necessary supporting agreements. This paper discusses the agreements and funding mechanisms used in support of the BCL shipments of TRU waste to Hanford for interim storage. In addition, this paper presents a summary of the efforts completed to demonstrate the effectiveness of the 10-160B Cask system. Lessons learned during this process are discussed and may be applicable to other TRU waste site shipment plans.« less

Novel Nuclear Powered Photocatalytic Energy Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

White,John R.; Kinsmen,Douglas; Regan,Thomas M.

2005-08-29

The University of Massachusetts Lowell Radiation Laboratory (UMLRL) is involved in a comprehensive project to investigate a unique radiation sensing and energy conversion technology with applications for in-situ monitoring of spent nuclear fuel (SNF) during cask transport and storage. The technology makes use of the gamma photons emitted from the SNF as an inherent power source for driving a GPS-class transceiver that has the ability to verify the position and contents of the SNF cask. The power conversion process, which converts the gamma photon energy into electrical power, is based on a variation of the successful dye-sensitized solar cell (DSSC)more » design developed by Konarka Technologies, Inc. (KTI). In particular, the focus of the current research is to make direct use of the high-energy gamma photons emitted from SNF, coupled with a scintillator material to convert some of the incident gamma photons into photons having wavelengths within the visible region of the electromagnetic spectrum. The high-energy gammas from the SNF will generate some power directly via Compton scattering and the photoelectric effect, and the generated visible photons output from the scintillator material can also be converted to electrical power in a manner similar to that of a standard solar cell. Upon successful implementation of an energy conversion device based on this new gammavoltaic principle, this inherent power source could then be utilized within SNF storage casks to drive a tamper-proof, low-power, electronic detection/security monitoring system for the spent fuel. The current project has addressed several aspects associated with this new energy conversion concept, including the development of a base conceptual design for an inherent gamma-induced power conversion unit for SNF monitoring, the characterization of the radiation environment that can be expected within a typical SNF storage system, the initial evaluation of Konarka's base solar cell design, the design and fabrication of a range of new cell materials and geometries at Konarka's manufacturing facilities, and the irradiation testing and evaluation of these new cell designs within the UML Radiation Laboratory. The primary focus of all this work was to establish the proof of concept of the basic gammavoltaic principle using a new class of dye-sensitized photon converter (DSPC) materials based on KTI's original DSSC design. In achieving this goal, this report clearly establishes the viability of the basic gammavoltaic energy conversion concept, yet it also identifies a set of challenges that must be met for practical implementation of this new technology.« less

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... of radioactive material under normal, off-normal, and credible accident conditions. (m) To the extent...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... of radioactive material under normal, off-normal, and credible accident conditions. (m) To the extent...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement... confinement of radioactive material under normal, off-normal, and credible accident conditions. (m) To the...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-02-01

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

A network of autism linked genes stabilizes two pools of synaptic GABAA receptors

PubMed Central

Tong, Xia-Jing; Hu, Zhitao; Liu, Yu; Anderson, Dorian; Kaplan, Joshua M

2015-01-01

Changing receptor abundance at synapses is an important mechanism for regulating synaptic strength. Synapses contain two pools of receptors, immobilized and diffusing receptors, both of which are confined to post-synaptic elements. Here we show that immobile and diffusing GABAA receptors are stabilized by distinct synaptic scaffolds at C. elegans neuromuscular junctions. Immobilized GABAA receptors are stabilized by binding to FRM-3/EPB4.1 and LIN-2A/CASK. Diffusing GABAA receptors are stabilized by the synaptic adhesion molecules Neurexin and Neuroligin. Inhibitory post-synaptic currents are eliminated in double mutants lacking both scaffolds. Neurexin, Neuroligin, and CASK mutations are all linked to Autism Spectrum Disorders (ASD). Our results suggest that these mutations may directly alter inhibitory transmission, which could contribute to the developmental and cognitive deficits observed in ASD. DOI: http://dx.doi.org/10.7554/eLife.09648.001 PMID:26575289

Bioremediation of Mercury by Vibrio fluvialis Screened from Industrial Effluents

PubMed Central

Saranya, Kailasam; Shekhar, Sudhanshu; Swaminathan, Sankaran; Balasubramanian, Thangavel

2017-01-01

Thirty-one mercury-resistant bacterial strains were isolated from the effluent discharge sites of the SIPCOT industrial area. Among them, only one strain (CASKS5) was selected for further investigation due to its high minimum inhibitory concentration of mercury and low antibiotic susceptibility. In accordance with 16S ribosomal RNA gene sequences, the strain CASKS5 was identified as Vibrio fluvialis. The mercury-removal capacity of V. fluvialis was analyzed at four different concentrations (100, 150, 200, and 250 μg/ml). Efficient bioremediation was observed at a level of 250 μg/ml with the removal of 60% of mercury ions. The interesting outcome of this study was that the strain V. fluvialis had a high bioremediation efficiency but had a low antibiotic resistance. Hence, V. fluvialis could be successfully used as a strain for the ecofriendly removal of mercury. PMID:28626761

Bioremediation of Mercury by Vibrio fluvialis Screened from Industrial Effluents.

PubMed

Saranya, Kailasam; Sundaramanickam, Arumugam; Shekhar, Sudhanshu; Swaminathan, Sankaran; Balasubramanian, Thangavel

2017-01-01

Thirty-one mercury-resistant bacterial strains were isolated from the effluent discharge sites of the SIPCOT industrial area. Among them, only one strain (CASKS5) was selected for further investigation due to its high minimum inhibitory concentration of mercury and low antibiotic susceptibility. In accordance with 16S ribosomal RNA gene sequences, the strain CASKS5 was identified as Vibrio fluvialis . The mercury-removal capacity of V. fluvialis was analyzed at four different concentrations (100, 150, 200, and 250  μ g/ml). Efficient bioremediation was observed at a level of 250  μ g/ml with the removal of 60% of mercury ions. The interesting outcome of this study was that the strain V. fluvialis had a high bioremediation efficiency but had a low antibiotic resistance. Hence, V. fluvialis could be successfully used as a strain for the ecofriendly removal of mercury.

Methods for Probabilistic Radiological Dose Assessment at a High-Level Radioactive Waste Repository.

NASA Astrophysics Data System (ADS)

Maheras, Steven James

Methods were developed to assess and evaluate the uncertainty in offsite and onsite radiological dose at a high-level radioactive waste repository to show reasonable assurance that compliance with applicable regulatory requirements will be achieved. Uncertainty in offsite dose was assessed by employing a stochastic precode in conjunction with Monte Carlo simulation using an offsite radiological dose assessment code. Uncertainty in onsite dose was assessed by employing a discrete-event simulation model of repository operations in conjunction with an occupational radiological dose assessment model. Complementary cumulative distribution functions of offsite and onsite dose were used to illustrate reasonable assurance. Offsite dose analyses were performed for iodine -129, cesium-137, strontium-90, and plutonium-239. Complementary cumulative distribution functions of offsite dose were constructed; offsite dose was lognormally distributed with a two order of magnitude range. However, plutonium-239 results were not lognormally distributed and exhibited less than one order of magnitude range. Onsite dose analyses were performed for the preliminary inspection, receiving and handling, and the underground areas of the repository. Complementary cumulative distribution functions of onsite dose were constructed and exhibited less than one order of magnitude range. A preliminary sensitivity analysis of the receiving and handling areas was conducted using a regression metamodel. Sensitivity coefficients and partial correlation coefficients were used as measures of sensitivity. Model output was most sensitive to parameters related to cask handling operations. Model output showed little sensitivity to parameters related to cask inspections.

Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, J.C.

2001-09-28

The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staffmore » has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they do demonstrate that the effect of BPRs is generally well behaved and that independent codes and cross-section libraries predict similar results. The report concludes with a discussion of the issues for consideration and recommendations for inclusion of SNF assemblies exposed to BPRs in criticality safety analyses using burnup credit for dry cask storage and transport.« less

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCann, R.A.; Lowery, P.S.; Lessor, D.L.

1987-09-01

HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations formore » conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs.« less

Test plan for evaluating the operational performance of the prototype nested, fixed-depth fluidic sampler

DOE Office of Scientific and Technical Information (OSTI.GOV)

REICH, F.R.

The PHMC will provide Low Activity Wastes (LAW) tank wastes for final treatment by a privatization contractor from two double-shell feed tanks, 241-AP-102 and 241-AP-104. Concerns about the inability of the baseline ''grab'' sampling to provide large volume samples within time constraints has led to the development of a nested, fixed-depth sampling system. This sampling system will provide large volume, representative samples without the environmental, radiation exposure, and sample volume impacts of the current base-line ''grab'' sampling method. A plan has been developed for the cold testing of this nested, fixed-depth sampling system with simulant materials. The sampling system willmore » fill the 500-ml bottles and provide inner packaging to interface with the Hanford Sites cask shipping systems (PAS-1 and/or ''safe-send''). The sampling system will provide a waste stream that will be used for on-line, real-time measurements with an at-tank analysis system. The cold tests evaluate the performance and ability to provide samples that are representative of the tanks' content within a 95 percent confidence interval, to sample while mixing pumps are operating, to provide large sample volumes (1-15 liters) within a short time interval, to sample supernatant wastes with over 25 wt% solids content, to recover from precipitation- and settling-based plugging, and the potential to operate over the 20-year expected time span of the privatization contract.« less

Temporal and Spatial Distribution of the Acetic Acid Bacterium Communities throughout the Wooden Casks Used for the Fermentation and Maturation of Lambic Beer Underlines Their Functional Role.

PubMed

De Roos, J; Verce, M; Aerts, M; Vandamme, P; De Vuyst, L

2018-04-01

Few data have been published on the occurrence and functional role of acetic acid bacteria (AAB) in lambic beer production processes, mainly due to their difficult recovery and possibly unknown role. Therefore, a novel aseptic sampling method, spanning both the spatial and temporal distributions of the AAB and their substrates and metabolites, was combined with a highly selective medium and matrix-assisted laser desorption ionization-time of flight mass spectrometry (MALDI-TOF MS) as a high-throughput dereplication method followed by comparative gene sequencing for their isolation and identification, respectively. The AAB ( Acetobacter species more than Gluconobacter species) proliferated during two phases of the lambic beer production process, represented by Acetobacter orientalis during a few days in the beginning of the fermentation and Acetobacter pasteurianus from 7 weeks until 24 months of maturation. Competitive exclusion tests combined with comparative genomic analysis of all genomes of strains of both species available disclosed possible reasons for this successive dominance. The spatial analysis revealed that significantly higher concentrations of acetic acid (from ethanol) and acetoin (from lactic acid) were produced at the tops of the casks, due to higher AAB counts and a higher metabolic activity of the AAB species at the air/liquid interface during the first 6 months of lambic beer production. In contrast, no differences in AAB species diversity occurred throughout the casks. IMPORTANCE Lambic beer is an acidic beer that is the result of a spontaneous fermentation and maturation process. Acidic beers are currently attracting attention worldwide. Part of the acidity of these beers is caused by acetic acid bacteria (AAB). However, due to their difficult recovery, they were never investigated extensively regarding their occurrence, species diversity, and functional role in lambic beer production. In the present study, a framework was developed for their isolation and identification using a novel aseptic sampling method in combination with matrix-assisted laser desorption ionization-time of flight mass spectrometry as a high-throughput dereplication technique followed by accurate molecular identification. The sampling method applied enabled us to take spatial differences into account regarding both enumerations and metabolite production. In this way, it was shown that more AAB were present and more acetic acid was produced at the air/liquid interface during a major part of the lambic beer production process. Also, two different AAB species were encountered, namely, Acetobacter orientalis at the beginning and Acetobacter pasteurianus in a later stage of the production process. This developed framework could also be applied for other fermentation processes. Copyright © 2018 American Society for Microbiology.

HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCann, R.A.; Lowery, P.S.

1987-10-01

HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equationsmore » for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs.« less

324 Building spent fuel segments pieces and fragments removal summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

SMITH, C L

2003-01-09

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

A&M. Radioactive parts security storage area. camera facing northwest. Outdoor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

A&M. Radioactive parts security storage area. camera facing northwest. Outdoor storage of concrete storage casks. Photographer: M. Holmes. Date: November 21, 1959. INEEL negative no. 59-6081 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

U.S. regulatory research program for implementation of burnup credit in transport casks

DOT National Transportation Integrated Search

2001-09-10

In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to : support the development of technical bases and guidance that would facilitate the implementation of : burnup credit into licensing activities for transport an...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2013 CFR

2013-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2014 CFR

2014-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2012 CFR

2012-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

Report on UQ and PCMM Analysis of Vacuum Drying for UFD S&T Gaps

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Fluss

2015-08-31

This report discusses two phenomena that could affect the safety, licensing, transportation, storage, and disposition of the spent fuel storage casks and their contents (radial hydriding during drying and water retention after drying) associated with the drying of canisters for dry spent fuel storage. The report discusses modeling frameworks and evaluations that are, or have been, developed as a means to better understand these phenomena. Where applicable, the report also discusses data needs and procedures for monitoring or evaluating the condition of storage containers during and after drying. A recommendation for the manufacturing of a fully passivated fuel rod, resistantmore » to oxidation and hydriding is outlined.« less

AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments weremore » transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.« less

Air Shipment of Spent Nuclear Fuel from Romania to Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Igor Bolshinsky; Ken Allen; Lucian Biro

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities formore » shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.« less

Alcohol promotions in Australian supermarket catalogues.

PubMed

Johnston, Robyn; Stafford, Julia; Pierce, Hannah; Daube, Mike

2017-07-01

In Australia, most alcohol is sold as packaged liquor from off-premises retailers, a market increasingly dominated by supermarket chains. Competition between retailers may encourage marketing approaches, for example, discounting, that evidence indicates contribute to alcohol-related harms. This research documented the nature and variety of promotional methods used by two major supermarket retailers to promote alcohol products in their supermarket catalogues. Weekly catalogues from the two largest Australian supermarket chains were reviewed for alcohol-related content over 12 months. Alcohol promotions were assessed for promotion type, product type, number of standard drinks, purchase price and price/standard drink. Each store catalogue included, on average, 13 alcohol promotions/week, with price-based promotions most common. Forty-five percent of promotions required the purchase of multiple alcohol items. Wine was the most frequently promoted product (44%), followed by beer (24%) and spirits (18%). Most (99%) wine cask (2-5 L container) promotions required multiple (two to three) casks to be purchased. The average number of standard drinks required to be purchased to participate in catalogue promotions was 31.7 (SD = 24.9; median = 23.1). The median price per standard drink was $1.49 (range $0.19-$9.81). Cask wines had the lowest cost per standard drink across all product types. Supermarket catalogues' emphasis on low prices/high volumes of alcohol reflects that retailers are taking advantage of limited restrictions on off-premise sales and promotion, which allow them to approach market competition in ways that may increase alcohol-related harms in consumers. Regulation of alcohol marketing should address retailer catalogue promotions. [Johnston R, Stafford J, Pierce H, Daube M. Alcohol promotions in Australian supermarket catalogues. Drug Alcohol Rev 2017;36:456-463]. © 2016 Australasian Professional Society on Alcohol and other Drugs.

Environmental data and analyses for the proposed management of spent nuclear fuel on the DOE Oak Ridge Reservation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Socolof, M.L.; Curtis, A.H.; Blasing, T.J.

1995-08-01

DOE needs to continue the safe and efficient management of SNF on ORR, based on the requirement for future SNF storage capacity and implementation of the ROD for the PEIS. DOE is proposing to implement the ROD through proper management of SNF on ORR, including the possible construction and operation of a dry cask storage facility. This report describes the potentially affected environment and analyzes impacts on various resources due to the proposed action. The information provided in this report is intended to support the Environmental Assessment being prepared for the proposed activities. Construction of the dry cask storage facilitymore » would result in minimal or no impacts on groundwater, surface water, and ecological resources. Contaminated soils excavated during construction would result in negligible risk to human health and to biota. Except for noise from trucks and equipment, operation of the dry cask storage facility would not be expected to have any impact on vegetation, wildlife, or rare plants or animals. Noise impacts would be minimal. Operation exposures to the average SNF storage facility worker would not exceed approximately 0.40 mSv/year (40 mrem/year). The off-site population dose within an 80-km (50-mile) radius of ORR from SNF operations would be less than 0.052 person-Sv/year (5.2 person-rem/year). Impacts from incident-free transportation on ORR would be less than 1.36 X 10{sup -4} occupational fatal cancers and 4.28 X 10{sup -6} public fatal cancers. Credible accident scenarios that would result in the greatest probable risks would cause less than one in a million cancer fatalities to workers and the public.« less

27 CFR 28.211 - General.

Code of Federal Regulations, 2010 CFR

2010-04-01

... TREASURY LIQUORS EXPORTATION OF ALCOHOL Exportation of Wine With Benefit of Drawback § 28.211 General. Wines manufactured, produced, bottled in bottles packed in containers, or packaged in casks or other... which are filled on premises qualified under this chapter to package or bottle wines, may, subject to...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stockman, Christine T.; Alsaed, Halim A.; Bryan, Charles R.

2015-03-26

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

Usage of FT-ICR-MS Metabolomics for Characterizing the Chemical Signatures of Barrel-Aged Whisky

PubMed Central

Roullier-Gall, Chloé; Signoret, Julie; Hemmler, Daniel; Witting, Michael A.; Kanawati, Basem; Schäfer, Bernhard; Gougeon, Régis D.; Schmitt-Kopplin, Philippe

2018-01-01

Whisky can be described as a complex matrix integrating the chemical history from the fermented cereals, the wooden barrels, the specific distillery processes, aging, and environmental factors. In this study, using Fourier transform ion cyclotron resonance mass spectrometry (FT-ICR-MS) and liquid chromatography coupled with tandem mass spectrometry (LC-MS/MS), we analyzed 150 whisky samples from 49 different distilleries, 7 countries, and ranging from 1 day new make spirit to 43 years of maturation with different types of barrel. Chemometrics revealed the unexpected impact of the wood history on the distillate's composition during barrel aging, regardless of the whisky origin. Flavonols, oligolignols, and fatty acids are examples of important chemical signatures for Bourbon casks, whereas a high number of polyphenol glycosides, including for instance quercetin-glucuronide or myricetin-glucoside as potential candidates, and carbohydrates would discriminate Sherry casks. However, the comparison of barrel aged rums and whiskies revealed specific signatures, highlighting the importance of the initial composition of the distillate and the distillery processes. PMID:29520358

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewes, J.

2014-02-24

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

Usage of FT-ICR-MS Metabolomics for characterizing the chemical signatures of barrel-aged whisky

NASA Astrophysics Data System (ADS)

Roullier-Gall, Chloé; Signoret, Julie; Hemmler, Daniel; Witting, Michael A.; Kanawati, Basem; Schäfer, Bernhard; Gougeon, Régis D.; Schmitt-Kopplin, Philippe

2018-02-01

Whisky can be described as a complex matrix integrating the chemical history from the fermented cereals, the wooden barrels, the specific distillery processes, ageing and environmental factors. In this study, using Fourier transform ion cyclotron resonance mass spectrometry (FT-ICR-MS) and liquid chromatography coupled with tandem mass spectrometry (LC-MS/MS), we analysed 150 whisky samples from 49 different distilleries, 7 countries, and ranging from 1 day new make spirit to 43 years of maturation with different types of barrel. Chemometrics revealed the unexpected impact of the wood history on the distillatés composition during barrel ageing, regardless of the whisky origin. Flavonols, oligolignols and fatty acids are examples of important chemical signatures for Bourbon casks, whereas a high number of polyphenol glycosides, including for instance quercetin-glucuronide or myricetin-glucoside as potential candidates, and carbohydrates would discriminate Sherry casks. However, the comparison of barrel aged rums and whiskies revealed specific signatures, highlighting the importance of the initial composition of the distillate and the distillery processes.

Apollo 12 Mission image - Modular Equipment Stowage Assemble (MESA) and the Fuel Cask on the Lunar Module (LM)

NASA Image and Video Library

1969-11-19

AS12-48-7034 (19 Nov. 1969) --- A close-up view of a portion of quadrant II of the descent stage of the Apollo 12 Lunar Module (LM), photographed during the Apollo 12 extravehicular activity (EVA). At lower left is the LM's Y footpad. The empty Radioisotope Thermoelectric Generator (RTG) fuel cask is at upper right. The fuel capsule has already been removed and placed in the RTG. The RTG furnishes power for the Apollo Lunar Surface Experiments Package (ALSEP) which the Apollo 12 astronauts deployed on the moon. The LM's descent engine skirt is in the center background. The rod-like object protruding out from under the footpad is a lunar surface sensing probe. Astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) in lunar orbit while astronauts Charles Conrad Jr., commander; and Alan L. Bean, lunar module pilot, descended in the LM to explore the moon.

SLUDGE TREATMENT PROJECT ENGINEERED CONTAINER RETRIEVAL AND TRANSFER SYSTEM PRELMINARY DESIGN HAZARD AND OPERABILITY STUDY

DOE Office of Scientific and Technical Information (OSTI.GOV)

CARRO CA

2011-07-15

This Hazard and Operability (HAZOP) study addresses the Sludge Treatment Project (STP) Engineered Container Retrieval and Transfer System (ECRTS) preliminary design for retrieving sludge from underwater engineered containers located in the 105-K West (KW) Basin, transferring the sludge as a sludge-water slurry (hereafter referred to as 'slurry') to a Sludge Transport and Storage Container (STSC) located in a Modified KW Basin Annex, and preparing the STSC for transport to T Plant using the Sludge Transport System (STS). There are six, underwater engineered containers located in the KW Basin that, at the time of sludge retrieval, will contain an estimated volumemore » of 5.2 m{sup 3} of KW Basin floor and pit sludge, 18.4 m{sup 3} of 105-K East (KE) Basin floor, pit, and canister sludge, and 3.5 m{sup 3} of settler tank sludge. The KE and KW Basin sludge consists of fuel corrosion products (including metallic uranium, and fission and activation products), small fuel fragments, iron and aluminum oxide, sand, dirt, operational debris, and biological debris. The settler tank sludge consists of sludge generated by the washing of KE and KW Basin fuel in the Primary Clean Machine. A detailed description of the origin of sludge and its chemical and physical characteristics can be found in HNF-41051, Preliminary STP Container and Settler Sludge Process System Description and Material Balance. In summary, the ECRTS retrieves sludge from the engineered containers and hydraulically transfers it as a slurry into an STSC positioned within a trailer-mounted STS cask located in a Modified KW Basin Annex. The slurry is allowed to settle within the STSC to concentrate the solids and clarify the supernate. After a prescribed settling period the supernate is decanted. The decanted supernate is filtered through a sand filter and returned to the basin. Subsequent batches of slurry are added to the STSC, settled, and excess supernate removed until the prescribed quantity of sludge is collected. The sand filter is then backwashed into the STSC. The STSC and STS cask are then inerted and transported to T Plant.« less

10 CFR 72.242 - Recordkeeping and reports.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Recordkeeping and reports. 72.242 Section 72.242 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT... Spent Fuel Storage Casks § 72.242 Recordkeeping and reports. (a) Each certificate holder or applicant...

The Role of the Neurofibromin-Syndecan-CASK Complex in the Regulation of Synaptic Ras-MAPK Signaling and Dendritic Spine Plasticity

DTIC Science & Technology

2006-02-01

Morgan, K., Hasz, D. E., Mao, Z., and Largaespada, D. A. (2005). Nf1 gene inactivation in acute myeloid leukemia cells confers cytarabine resistance through MAPK and mTOR pathways. Leukemia. VII. Appendices: None

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J.; Marshall, William B. J.; Ilas, Germina

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 guidancemore » 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.« less

27 CFR 25.35 - Tanks.

Code of Federal Regulations, 2010 CFR

2010-04-01

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

27 CFR 25.35 - Tanks.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2013-04-01 2013-04-01 false Tanks. 25.35 Section 25.35... TREASURY ALCOHOL BEER Construction and Equipment Equipment § 25.35 Tanks. Each stationary tank, vat, cask... contents of tanks or containers in lieu of providing each tank or container with a measuring device. (Sec...

10 CFR 72.234 - Conditions of approval.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Conditions of approval. 72.234 Section 72.234 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT... Spent Fuel Storage Casks § 72.234 Conditions of approval. (a) The certificate holder and applicant for a...

10 CFR 72.232 - Inspection and tests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Inspection and tests. 72.232 Section 72.232 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL... Storage Casks § 72.232 Inspection and tests. (a) The certificate holder and applicant for a CoC shall...

76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-17

.... FOR FURTHER INFORMATION CONTACT: Matthew Gordon, Structural Mechanics and Materials Branch, Division... a fee. Comments and questions on ISG-23 should be directed to Matthew Gordon, Structural Mechanics..., 2011. For the U.S. Nuclear Regulatory Commission. Michele Sampson, Acting Chief, Structural Mechanics...

27 CFR 25.35 - Tanks.

Code of Federal Regulations, 2012 CFR

2012-04-01

... TREASURY LIQUORS BEER Construction and Equipment Equipment § 25.35 Tanks. Each stationary tank, vat, cask or other container used, or intended for use, as a receptacle for wort, beer or concentrate produced from beer shall: (a) Be durably marked with a serial number and capacity; and (b) Be equipped with a...

27 CFR 25.35 - Tanks.

Code of Federal Regulations, 2014 CFR

2014-04-01

... TREASURY ALCOHOL BEER Construction and Equipment Equipment § 25.35 Tanks. Each stationary tank, vat, cask or other container used, or intended for use, as a receptacle for wort, beer or concentrate produced from beer shall: (a) Be durably marked with a serial number and capacity; and (b) Be equipped with a...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-16

.... The NRC is taking this action because the applicant identified that a certain Technical Specification (TS) for Boral characterization was not written precisely. Specifically, the requirements for meeting... changes to the technical specifications. The NRC also published a direct final rule on May 6, 2010 (75 FR...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-16

...) Number 1030. The NRC is taking this action because the applicant identified that a certain Technical Specification (TS) for Boral characterization was not written precisely and in a manner that could be readily... cavity water removal operations, and making [[Page 41370

78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-17

... requirement that loaded storage casks also meet transportation requirements. Integration of storage and... transported from the storage location. As part of its evaluation of integration and compatibility between... evaluating compatibility of storage and transportation regulations. As part of its evaluation of integration...

27 CFR 25.35 - Tanks.

Code of Federal Regulations, 2011 CFR

2011-04-01

... TREASURY LIQUORS BEER Construction and Equipment Equipment § 25.35 Tanks. Each stationary tank, vat, cask or other container used, or intended for use, as a receptacle for wort, beer or concentrate produced from beer shall: (a) Be durably marked with a serial number and capacity; and (b) Be equipped with a...

OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-09-01

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

78 FR 79021 - Sunshine Act Meeting Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-27

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0001] Sunshine Act Meeting Notice DATES: Weeks of December... meetings scheduled for the week of December 30, 2013. Week of January 6, 2014--Tentative Monday, January 6... to Dry Casks (Public Meeting) (Contact: Kevin Witt, 301-415-2145). This meeting will be webcast live...

27 CFR 26.40 - Marking containers of distilled spirits.

Code of Federal Regulations, 2010 CFR

2010-04-01

... spirits. The distiller, rectifier, or bottler shall serially number each case, barrel, cask, or similar... the container, the distiller, rectifier, or bottler shall plainly print, stamp, or stencil with..., rectifier, or bottler. (b) The brand name and kind of liquor; (c) The wine and proof gallon contents; or...

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pancake, Daniel C.; Rock, Cynthia

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory mademore » the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saltzstein, Sylvia J.; Sorenson, Ken B.; Hanson, Brady

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 andmore » 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.« less

Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cochran, John R.; Hardin, Ernest

2015-07-01

This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portionsmore » of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has benefited greatly from review principally by Steve Pye, and also by Paul Eslinger, Dave Sevougian and Jiann Su.« less

10 CFR 72.48 - Changes, tests, and experiments.

Code of Federal Regulations, 2011 CFR

2011-01-01

... facility or spent fuel storage cask design, of changes in procedures, and of tests and experiments made... 10 Energy 2 2011-01-01 2011-01-01 false Changes, tests, and experiments. 72.48 Section 72.48... Issuance and Conditions of License § 72.48 Changes, tests, and experiments. (a) Definitions for the...

10 CFR 72.48 - Changes, tests, and experiments.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

27 CFR 26.207 - Destruction of marks and brands.

Code of Federal Regulations, 2010 CFR

2010-04-01

... brands. 26.207 Section 26.207 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE... Products Coming Into the United States From the Virgin Islands § 26.207 Destruction of marks and brands. The marks, brands, and serial numbers required by this part to be placed on barrels, casks, or similar...

27 CFR 26.41 - Destruction of marks and brands.

Code of Federal Regulations, 2010 CFR

2010-04-01

... brands. 26.41 Section 26.41 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE... Products Coming Into the United States From Puerto Rico § 26.41 Destruction of marks and brands. The marks, brands, and serial numbers required by this part to be placed on barrels, casks, or similar containers...

9. DETAIL VIEW OF BRIDGE CRANE ON WEST SIDE OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. DETAIL VIEW OF BRIDGE CRANE ON WEST SIDE OF BUILDING. CAMERA FACING NORTHEAST. CONTAMINATED AIR FILTERS LOADED IN TRANSPORT CASKS WERE TRANSFERRED TO VEHICLES AND SENT TO RADIOACTIVE WASTE MANAGEMENT COMPLEX FOR STORAGE. INEEL PROOF NUMBER HD-17-1. - Idaho National Engineering Laboratory, Old Waste Calcining Facility, Scoville, Butte County, ID

78 FR 74188 - Sunshine Act Meetings Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-10

... Casks (Public Meeting) (Contact: Kevin Witt, 301-415-2145) This meeting will be Web cast live at the Web... Weather Events (Public Meeting) (Contact: George Wilson, 301-415-1711) This meeting will be Web cast live... will be Web cast live at the Web address-- http://www.nrc.gov/ . Week of January 13, 2014--Tentative...

Between wilderness and the middle landscape: A rocky road

Treesearch

Lisi Krall

2007-01-01

Wilderness preservation, as one branch of conservation, demonstrates a decidedly different cultural ethos than the utilitarian branch. Thus, preservation and utilitarian conservation represent different habits of thought fermenting in the cask of l9th century economic evolution. More specifically, the utilitarian branch of conservation can easily be viewed as an...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

..., and that are not in areas of known seismic activity, a standardized design earthquake ground motion... motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault... of the Design Earthquake Ground Motion (DE). The DE for the site is characterized by both horizontal...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

..., and that are not in areas of known seismic activity, a standardized design earthquake ground motion... motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault... of the Design Earthquake Ground Motion (DE). The DE for the site is characterized by both horizontal...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

..., and that are not in areas of known seismic activity, a standardized design earthquake ground motion... motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault... of the Design Earthquake Ground Motion (DE). The DE for the site is characterized by both horizontal...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

..., and that are not in areas of known seismic activity, a standardized design earthquake ground motion... motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault... of the Design Earthquake Ground Motion (DE). The DE for the site is characterized by both horizontal...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

..., and that are not in areas of known seismic activity, a standardized design earthquake ground motion... motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault... of the Design Earthquake Ground Motion (DE). The DE for the site is characterized by both horizontal...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finch, Robert J.; Smartt, Heidi A.; Haddal, Risa

Once a geological repository has begun operations, the encapsulation and disposal of spent fuel will be performed as a continuous, industrial-scale series of processes, during which time safeguards seals will be applied to transportation casks before shipment from an encapsulation plant, and then verified and removed following receipt at the repository. These operations will occur approximately daily during several decades of Sweden's repository operation; however, requiring safeguards inspectors to perform the application, verification, and removal of every seal would be an onerous burden on International Atomic Energy Agency's (IAEA's) resources. Current IAEA practice includes allowing operators to either apply sealsmore » or remove them, but not both, so the daily task of either applying or verifying and removing would still require continuous presence of IAEA inspectors at one site at least. Of special importance is the inability to re-verify cask or canisters from which seals have been removed and the canisters emplaced underground. Successfully designing seals that can be applied, verified and removed by an operator with IAEA approval could impact more than repository shipments, but other applications as well, potentially reducing inspector burdens for a wide range of such duties.« less

Status of a standard for neutron skyshine calculation and measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westfall, R.M.; Wright, R.Q.; Greenborg, J.

1990-01-01

An effort has been under way for several years to prepare a draft standard, ANS-6.6.2, Calculation and Measurement of Direct and Scattered Neutron Radiation from Contained Sources Due to Nuclear Power Operations. At the outset, the work group adopted a three-phase study involving one-dimensional analyses, a measurements program, and multi-dimensional analyses. Of particular interest are the neutron radiation levels associated with dry-fuel storage at reactor sites. The need for dry storage has been investigated for various scenarios of repository and monitored retrievable storage (MRS) facilities availability with the waste stream analysis model. The concern is with long-term integrated, low-level dosesmore » at long distances from a multiplicity of sources. To evaluate the conservatism associated with one-dimensional analyses, the work group has specified a series of simple problems. Sources as a function of fuel exposure were determined for a Westinghouse 17 x 17 pressurized water reactor assembly with the ORIGEN-S module of the SCALE system. The energy degradation of the 35 GWd/ton U sources was determined for two generic designs of dry-fuel storage casks.« less

Analysis of the factors that impact the reliability of high level waste canister materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyd, W.K.; Hall, A.M.

1977-09-19

The analysis encompassed identification and analysis of potential threats to canister integrity arising in the course of waste solidification, interim storage at the fuels reprocessing plant, wet and dry shipment, and geologic storage. Fabrication techniques and quality assurance requirements necessary to insure optimum canister reliability were considered taking into account such factors as welding procedure, surface preparation, stress relief, remote weld closure, and inspection methods. Alternative canister materials and canister systems were also considered in terms of optimum reliability in the face of threats to the canister's integrity, ease of fabrication, inspection, handling and cost. If interim storage in airmore » is admissible, the sequence suggested comprises producing a glass-type waste product in a continuous ceramic melter, pouring into a carbon steel or low-alloy steel canister of moderately heavy wall thickness, storing in air upright on a pad and surrounded by a concrete radiation shield, and thereafter placing in geologic storage without overpacking. Should the decision be to store in water during the interim period, then use of either a 304 L stainless steel canister overpacked with a solution-annealed and fast-cooled 304 L container, or a single high-alloy canister, is suggested. The high alloy may be Inconel 600, Incoloy Alloy 800, or Incoloy Alloy 825. In either case, it is suggested that the container be overpacked with a moderately heavy wall carbon steel or low-alloy steel cask for geologic storage to ensure ready retrievability. 19 figs., 5 tables.« less

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE PAGES

Jiang, Hao; Wang, Jy-An John

2016-10-14

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Parallelization of KENO-Va Monte Carlo code

NASA Astrophysics Data System (ADS)

Ramón, Javier; Peña, Jorge

1995-07-01

KENO-Va is a code integrated within the SCALE system developed by Oak Ridge that solves the transport equation through the Monte Carlo Method. It is being used at the Consejo de Seguridad Nuclear (CSN) to perform criticality calculations for fuel storage pools and shipping casks. Two parallel versions of the code: one for shared memory machines and other for distributed memory systems using the message-passing interface PVM have been generated. In both versions the neutrons of each generation are tracked in parallel. In order to preserve the reproducibility of the results in both versions, advanced seeds for random numbers were used. The CONVEX C3440 with four processors and shared memory at CSN was used to implement the shared memory version. A FDDI network of 6 HP9000/735 was employed to implement the message-passing version using proprietary PVM. The speedup obtained was 3.6 in both cases.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Bias estimates used in lieu of validation of fission products and minor actinides in MCNP K eff calculations for PWR burnup credit casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, Don E.; Marshall, William J.; Wagner, John C.

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 (k eff) 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 biasmore » due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff 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.« less

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

Code of Federal Regulations, 2010 CFR

2010-04-01

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

ROUGHING IT. THE OLD MAN AND THE SEA. SHORT STORIES. POEMS. LITERATURE CURRICULUM III, TEACHER VERSION.

ERIC Educational Resources Information Center

KITZHABER, ALBERT R.

A TEACHER VERSION OF A LITERATURE CURRICULUM GUIDE WAS PROVIDED FOR TWAIN'S "ROUGHING IT," HEMINGWAY'S "THE OLD MAN AND THE SEA," FOUR SHORT STORIES, AND 20 LYRIC POEMS. THE SHORT STORIES INCLUDED WERE (1) "THE MONKEY'S PAW" BY W.W. JACOBS, (2) "PAUL'S CASE" BY WILLA CATHER, (3) "THE CASK OF…

Heat transfer to four fineness-ratio-1.6 hexagonal prisms with various corner radii at Mach 6

NASA Technical Reports Server (NTRS)

Hunt, J. L.

1972-01-01

An investigation was conducted in the Langley 20-inch Mach 6 tunnel to define the aerodynamic heat transfer to the radioisotope fuel cask (heat source) of the SNAP-19/Pioneer power system. The shape of the SNAP-19/Pioneer heat source is that of a hexagonal prism with flat ends; the fineness ratio, based on maximum (edge to edge) diameter, is 1.61. Phase-change-paint heat-transfer data and schlieren photographs were obtained on four possible 1/2-scale entry configurations of the SNAP-19/Pioneer heat source. Tests were conducted over a wide range of attitudes and at nominal Reynolds numbers, based on the length of the unablated configuration, of 33,000; 84,000; and 2,200,000.

Ultrasonic fingerprinting by phased array transducer

NASA Astrophysics Data System (ADS)

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

2016-06-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meyer, Ryan M.; Suter, Jonathan D.; Jones, Anthony M.

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.

ROUGHING IT. THE OLD MAN AND THE SEA. SHORT STORIES. LYRIC POETRY. LITERATURE CURRICULUM III, STUDENT VERSION.

ERIC Educational Resources Information Center

KITZHABER, ALBERT R.

A STUDENT VERSION OF A LITERATURE CURRICULUM GUIDE WAS PROVIDED FOR TWAIN'S "ROUGHING IT," HEMINGWAY'S "THE OLD MAN AND THE SEA," FOUR SHORT STORIES, AND 20 LYRIC POEMS. THE SHORT STORIES INCLUDED WERE (1) "THE MONKEY'S PAW" BY W.W. JACOBS, (2) "PAUL'S CASE" BY WILLA CATHER, (3) "THE CASK OF…

75 FR 23820 - Notice of Docketing of Amendment Request for Materials License No. SNM-2506; Northern States...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-04

... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... Generating Plant (PINGP), Unit Nos. 1 and 2, site in Goodhue County, Minnesota. The TN-40 cask is currently..., higher burnup spent fuel used in the PINGP reactor as well as associated changes to the ISFSI's technical...

Constructing Complexity: Using Reading Levels to Differentiate Reading Comprehension Activities

ERIC Educational Resources Information Center

FitzPatrick, Declan

2008-01-01

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

SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allender, Jeffrey S.; Bridges, Nicholas J.; Loftin, Bradley M.

2015-07-14

Savannah River National Laboratory (SRNL) and Oak Ridge National Laboratory (ORNL) are developing plans for the recovery of rare and unique isotopes contained within heavy-actinide target assemblies, specifically the Mark-18A. Mark-18A assemblies were irradiated in Savannah River Site (SRS) reactors in the 1970s under extremely high neutron-flux conditions and produced, virtually, the world's supply of plutonium-244, an isotope of key importance to high-precision actinide measurement and other scientific and nonproliferation uses; and curium highly enriched in heavy isotopes (e.g., curium-246 and curium-248). In 2015 and 2016, SRNL is pursuing tasks that would reduce program risk and budget requirements, including furthermore » characterization of unprocessed targets; engineering studies for the use of the SRNL Shielded Cells Facility (SCF) for recovery; and development of onsite and offsite shipping methods including a replacement for the heavy (70 ton) cask previously used for onsite transfer of irradiated items at SRS. A status update is provided for the characterization, including modeling using the Monte Carlo N-Particle Transport Code (MCNP); direct non-destructive assay measurements; and cask design.« less

FRAPCON analysis of cladding performance during dry storage operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richmond, David J.; Geelhood, Kenneth J.

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservativelymore » showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.« less

Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

KESSLER, S.F.

This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weldmore » station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.« less

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlsen, Brett; Chiguer, Mustapha; Grahn, Per

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years untilmore » reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.« less

Licensing a new industrial irradiator.

PubMed

Bates, Nicolas K; Entwistle, Frederick B

2010-02-01

After nearly three decades of medical product sterilization, 3M launched a major new project to build and license an irradiator facility. 3M Corporate Health Physics was responsible for the licensing aspect of this project. The licensing process consisted of six amendments, over 30 submissions to the U.S. Nuclear Regulatory Commission (U.S. NRC) and four U.S. NRC site visits. It took approximately 22 months to complete. The six license amendments are reviewed and several of the submissions are discussed. These include 3M's response to the U.S. NRC's interest in the shielding calculations used for the bioshield, the development of a protocol of radiation safety system test methods, and an analysis to show that a dropped cask during loading operations would not fall on sealed sources. A number of lessons were learned during the course of licensing the new irradiator. Among these were the importance of understanding the U.S. NRC license reviewer's perspective, the need to thoroughly review the irradiator manufacturer's licensing package during project negotiations, the benefits of leaving the Health Physics Office and meeting with the non-health physicists involved in the project, and the necessity of maintaining the solid relationships that already existed with the site Radiation Safety Officer and Sterilization Engineer.

Planning and supervision of reactor defueling using discrete event techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garcia, H.E.; Imel, G.R.; Houshyar, A.

1995-12-31

New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on themore » progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.« less

Update Direct-Strike Lightning Environment for Stockpile-to-Target Sequence: Supplement LLNL Subcontract #B568621 Lightning Protection at the Yucca Mountain Waste Storage Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uman, M A

2008-10-09

The University of Florida has surveyed all relevant publications reporting lightning damage to metals, metals which could be used as components of storage containers for nuclear waste materials. We show that even the most severe lightning could not penetrate the stainless steel thicknesses proposed for nuclear waste storage casks.

Physics Flash August 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kippen, Karen Elizabeth

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."

19 CFR 4.7 - Inward foreign manifest; production on demand; contents and form; advance filing of cargo...

Code of Federal Regulations, 2012 CFR

2012-04-01

... enclosed in any container such as a box, bale, bag, cask, or the like. Such cargo is also described as bulk...; contents and form; advance filing of cargo declaration. 4.7 Section 4.7 Customs Duties U.S. CUSTOMS AND... and form; advance filing of cargo declaration. (a) The master of every vessel arriving in the United...

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2012-06-26

145 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial ...Pakistan’s Civil Nuclear Program.” Some analysts argue that spent nuclear fuel is more vulnerable when being transported . 146 Martellini, 2008. 147...produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment technology, which it mastered by the mid-1980s

Nuclear Energy Policy

DTIC Science & Technology

2008-01-28

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

DOE PAGES

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

2018-02-26

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« less

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« less

A&M. TAN633. Sections show view of hot cell caskentry doors, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

A&M. TAN-633. Sections show view of hot cell cask-entry doors, manipulators in each cell, drainage trenches, door and room details. Ralph M. Parsons 1229-13-ANP/GE-3-633-A-2. Date: December 1956. Approved by INEEL Classification Office for public release. INNEL index code no. 034-0633-00-693-107316 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-07-30

Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 79...that Pakistan’s strategic nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs...that gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2010-10-07

Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 99...prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage facilities and personnel reliability... nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs of Staff Admiral Michael

Review and Implementation of Technology for Solid Radioactive Waste Volume Reduction

DTIC Science & Technology

1999-10-15

were shifted to Project 1.1 for spent nuclear fuel cask development to accelerate that project. Those funds should be repaid to Project 1.3 in the... transported between the shipyards such as Nerpa, and other intermediate storage sites such as Gremikha and Andreeva Bay. At these sites the largest...waste source and allow pretreatment unit operations using commercially available technologies of contaminant assaying, cutting/shearing, sorting

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-10-15

and technical measures to prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage ...Talks On Nuclear Security,” The Boston Globe, May 5, 2009. 79 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or...a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 80 Martellini, 2008. 81 For more information

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2012-05-10

2009. 143 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in...Program.” Some analysts argue that spent nuclear fuel is more vulnerable when being transported . 144 Martellini, 2008. Pakistan’s Nuclear Weapons...urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment technology, which it

Technological and microbiological aspects of traditional balsamic vinegar and their influence on quality and sensorial properties.

PubMed

Giudici, Paolo; Gullo, Maria; Solieri, Lisa; Falcone, Pasquale Massimiliano

2009-01-01

The term "balsamic" is widespread and popular all over the world of vinegar and fancy foods; it is used generally to refer to vinegars and sauces with a sweet and sour taste. However, the original is the European Protected Denomination, registered as "Aceto Balsamico Tradizionale of Modena, or of Reggio Emilia" that should not be confused with the "Aceto Balsamico di Modena" very similar in the name, but completely different for technology, raw material, quality, and sensorial properties. Traditional balsamic vinegar is made by a peculiar procedure, that starts with a thermal concentration of freshly squeezed grape juice, followed by alcoholic and acetic fermentations and, finally, long aging in a wooden barrel set, by a procedure which requires a partial transfer of vinegar from cask to cask with the consequential blending of vinegars of different ages. In addition, water transfer occurs across the wood of the barrels, the result being an increase of solute concentration of the vinegar. The chemical and physical transformations of the vinegar are mainly directed by the low water activity of the vinegar. High-molecular polymeric compounds are the main and characteristic constituents of original and old traditional balsamic vinegar, and the major cause of its rheological and sensorial properties.

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregson, Michael Warren; Mo, Tin; Sorenson, Ken Bryce

The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratoriesmore » has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.« less

Projected Standard on neutron skyshine. [Skyshine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westfall, R.M.; Williams, D.S.

1987-07-01

Current interest in neutron skyshine arises from the application of dry fuel handling and storage techniques at reactor sites, at the proposed monitored retrievable storage facility and at other facilities being considered as part of the civilian radioactive waste management programs. The chairman of Standards Subcommittee ANS-6, Radiation Protection and Shielding, has requested that a work group be formed to characterize the neutron skyshine problem and, if necessary, prepare a draft Standard. The work group is comprised of representatives of storage cask vendors, architect engineering firms, nuclear utilities, the academic community and staff members of national laboratories and government agencies.more » The purpose of this presentation summary is to describe the activities of the work group and the scope and contents of the projected Standard, ANS-6.6.2, ''Calculation and Measurement of Direct and Scattered Neutron Radiation from Nuclear Power Operations.'' The specific source under consideration by the work group is an array of dry fuel casks located at a reactor site. However, it is recognized that the scope of the standard should be broad enough to encompass other neutron sources. The Standard will define appropriate methodology for properly characterizing the neutron dose due to skyshine. This dose characterization is necessary, for example, in demonstrating compliance with pertinent regulatory criteria.« less

HAZARDS OF THERMAL EXPANSION FOR RADIOLOGICAL CONTAINER ENGULFED IN FIRE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donna Post Guillen

2013-05-01

Fire accidents pose a serious threat to nuclear facilities. It is imperative that transport casks or shielded containers designed to transport/contain radiological materials have the ability to withstand a hypothetical fire. A numerical simulation was performed for a shielded container constructed of stainless steel and lead engulfed in a hypothetical fire as outlined by 10 CFR §71.73. The purpose of this analysis was to determine the thermal response of the container during and after the fire. The thermal model shows that after 30 minutes of fire, the stainless steel will maintain its integrity and not melt. However, the lead shieldingmore » will melt since its temperature exceeds the melting point. Due to the method of construction of the container under consideration, ample void space must be provided to allow for thermal expansion of the lead upon heating and melting, so as to not overstress the weldment.« less

Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.

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 (UNF) and high-level radioactive waste (HLW). 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 and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extendedmore » storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact those loads may have on the integrity of the UNF assembly. The shock and vibration testing along with computer modeling is a vital part of research to achieve closure of a gap in information related to the ability of ES UNF to maintain its safety function when subjected to NCT. In support of this effort, preliminary structural dynamics modeling is presented herein. The modeling investigates the rigidity of a hypothetical cask and cradle structure by comparing it to a monolithic concrete mass. The concrete mass represents a practical option for achieving the necessary cask and cradle mass on a flatbed railcar, but this comparative modeling study investigates whether or not the dynamic loads transmitted through a monolithic concrete configuration are adequately representative of a realistic cask and cradle system. This modeling highlights the need for rail testing by reporting the phenomenon of structural transmissibility. As shown herein, this structural transmissibility can cause an amplification of shock and vibration loads through the structure, which could potentially lead to accelerated mechanical degradation of UNF under NCT.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sobes, Vladimir; Scaglione, John M; Wagner, John C

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 allowmore » 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 k eff). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sobes, Vladimir; Scaglione, John M.; Wagner, John C.

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 allowmore » 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 k eff). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.« less

Transportation accident scenarios for commercial spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilmot, E L

1981-02-01

A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

Transportation of spent MTR fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raisonnier, D.

1997-08-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryan, Charles R.; Enos, David

2014-09-01

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

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2010-02-04

Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008...measures to prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage facilities and personnel...strategic nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs of Staff Admiral

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-12-09

Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008...gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment...technology, which it mastered by the mid-1980s. Highly-enriched uranium (HEU) is one of two types of fissile material used in nuclear weapons; the other

Environmental Assessment for Enhanced Use Leasing West Side Development, Phase I South, Hill AFB, Utah

DTIC Science & Technology

2006-09-01

training speeds into one or several of hundreds of nuclear fuel rod storage casks could release immensely toxic radioactive wastes that have a 10,000...distinctions between the risks related to open storage of spent nuclear fuel rods in Skull Valley and the risks to civilian facilities within the...operations, stores, markets, coffee shops and other strictly civilian commercial enterprises. No family or residential housing use is proposed

High Fragmentation Steel Production Process

DTIC Science & Technology

1984-01-01

J/ FTA c« ;« MO G SO KM s s P WS W-U Hi ; T 14 434 CASK G S3 K 11 ma WM MM MM ACTS 1 TC*4 U S7« ill GC 135 V M NTA «M FT...relative feed range 2nd digit -relative force range FMd 1 Very Low Fore* t 2 Low 2 3 Medium Low 3 4 Medium 4 5 Medium 5 6 Medium High 6 7 Medium

Process and equipment development for hot isostatic pressing treatability study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bateman, Ken; Wahlquist, Dennis; Malewitz, Tim

2015-03-01

Battelle Energy Alliance (BEA), LLC, has developed processes and equipment for a pilot-scale hot isostatic pressing (HIP) treatability study to stabilize and volume reduce radioactive calcine stored at Idaho National Laboratory (INL). In 2009, the U. S. Department of Energy signed a Record of Decision with the state of Idaho selecting HIP technology as the method to treat 5,800 yd^3 (4,400 m^3) of granular zirconia and alumina calcine produced between 1953 and 1992 as a waste byproduct of spent nuclear fuel reprocessing. Since the 1990s, a variety of radioactive and hazardous waste forms have been remotely treated using HIP withinmore » INL hot cells. To execute the remote process at INL, waste is loaded into a stainless-steel or aluminum can, which is evacuated, sealed, and placed into a HIP furnace. The HIP simultaneously heats and pressurizes the waste, reducing its volume and increasing its durability. Two 1 gal cans of calcine waste currently stored in a shielded cask were identified as candidate materials for a treatability study involving the HIP process. Equipment and materials for cask-handling and calcine transfer into INL hot cells, as well as remotely operated equipment for waste can opening, particle sizing, material blending, and HIP can loading have been designed and successfully tested. These results demonstrate BEA’s readiness for treatment of INL calcine.« less

Safety aspects of nuclear waste disposal in space

NASA Technical Reports Server (NTRS)

Rice, E. E.; Edgecombe, D. S.; Compton, P. R.

1981-01-01

Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

Unique Chernobyl Cranes for Deconstruction Activities in the New Safe Confinement - 13542

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parameswaran, N.A. Vijay; Chornyy, Igor; Owen, Rob

2013-07-01

The devastation left behind from the Chernobyl nuclear power plant (ChNPP) Unit 4 accident which occurred on April 26, 1986 presented unparalleled technical challenges to the world engineering and scientific community. One of the largest tasks that are in progress is the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter (OS) that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant (ChNPP) Unit. One of the major mechanical handling systems to be installed in themore » NSC is the Main Cranes System (MCS). The planned decontamination and decommissioning or dismantling (D and D) activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the NSC, will require large and sophisticated cranes. The article will focus on the unique design features of the MCS for the D and D activities. (authors)« less

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enercon Services, Inc.

2011-03-14

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnupmore » Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.« less

Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cherkas, Dmytro

2011-10-01

As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, andmore » lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the ‘add a source’ basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.« less

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

2011-01-01

The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address themore » issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias and uncertainty results based on a quality-assurance-controlled prerelease version of the Scale 6.1 code package and the ENDF/B-VII nuclear cross section data.« less

Job Samples as Tank Gunnery Performance Predictors

DTIC Science & Technology

1980-09-01

previously been obtained ( Maitland , Eaton, and Neff, 1980). Table VI-M. The Table VI-M order of firing and engagement techniques used in Phase III were the... Maitland , Eaton, and Neff (1980) *** <.001 22 Sen 21 ’ )a. datz obtained from ch4 sensing cask was quanti:ied. as in Phase It, by Coti.purtirg the vie.i...of four test weighting methods in multiple regression. Educational and Psychological Measure- ment, 1959, 19, 103-114. Maitland , A. J., Eaton, N. K

Index to FAA Office of Aviation Medicine Reports: 1961 through 1980,

DTIC Science & Technology

1981-01-01

in ballistocardiographic research and the current state of the art . AD455651 64-13 Gogel, W. C.: The size cue to visually perceived distance...AD773451 73-12 Lewis, M. F., and Ferraro, D. P.: Flying high: The aeromedical aspects of marihuana . AD775889 73-13 Tobias, J. V., and Irons, F. M...Ferraro, D. P., Mertens, H. W., and Steen, J. A.: Interaction between marihuana and altitude on a complex behavioral cask in baboons. ADA020680/5GI 75

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Hu, Jianwei; De Baere, P.

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 frameworkmore » 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 verification technique.« less

Bounding criticality safety analyses for shipments of unconfigured spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lichtenwalter, J.J.; Parks, C.V.

1998-06-01

In November 1996, a request was made to the US Department of Energy for a waiver for three shipments of spent nuclear fuel (SNF) from Oak Ridge National Laboratory (ORNL) to the Savannah River Site (SRS) in the US NRC certified BMI-1 cask (CoC 5957). Although the post-irradiation fissile mass (based on chemical assays) in each shipment was less than 800 g, a criticality safety analysis was needed because the pre-irradiation mass exceeded 800 g, the fissile material limit in the CoC. The analyses were performed on SNF consisting of aluminum-clad U{sub 3}O{sub 8}, UAl{sub x}, and U{sub 3}Si{sub 2}more » plates, fragments and pieces that had been irradiated at ORNL during the Reduced Enrichment Research and Test Reactor Program of the 1980s. The highlights of the approach used to analyze this unique SNF and the benefits of the waiver are presented in this paper.« less

Descriptive sensory analysis of Aceto Balsamico Tradizionale di Modena DOP and Aceto Balsamico Tradizionale di Reggio Emilia DOP.

PubMed

Zeppa, Giuseppe; Gambigliani Zoccoli, Mario; Nasi, Enrico; Masini, Giovanni; Meglioli, Giuseppe; Zappino, Matteo

2013-12-01

Aceto Balsamico Tradizionale (ABT) is a typical Italian vinegar available in two different forms: Aceto Balsamico Tradizionale di Modena DOP (ABTM) and Aceto Balsamico Tradizionale di Reggio Emilia DOP (ABTRE). ABT is obtained by alcoholic fermentation and acetic bio-oxidation of cooked grape must and aged at least 12 years in wooden casks and is known and sold around the world. Despite this widespread recognition, data on sensory characteristics of these products are very scarce. Therefore a descriptive analysis was conducted to define a lexicon for the ABT sensory profile and to create a simple, stable and reproducible synthetic ABT for training panellists. A lexicon of 20 sensory parameters was defined and validated and a synthetic ABT was prepared as standard reference. Simple standards for panellist training were also defined and the sensory profiles of ABTM and ABTRE were obtained. The obtained results confirm that descriptive analysis can be used for the sensory characterisation of ABT and that the sensory profiles of ABTM and ABTRE are very different. Furthermore, the results demonstrate that a lexicon and proper standard references are essential for describing the sensory qualities of ABT both for technical purposes and to protect the product from commercial fraud. © 2013 Society of Chemical Industry.

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

Systems for the Intermodal Routing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom

The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encounteredmore » during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M; Montgomery, Rose; Bevard, Bruce Balkcom

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.

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Cask weeping mitigation

DOEpatents

Krumhansl, James L.; Brady, Patrick V.; Teter, David M.; McConnell, Paul

2007-09-18

A method (and concomitant kit) for treating a surface to reduce subsequent .sup.137Cs nuclide desorption comprising contacting the surface with a first cation-containing solution, the cation being one or more of Cs.sup.+, Rb.sup.+, Ag.sup.+, Tl.sup.+, K.sup.+, and NH.sub.4.sup.+, and contacting the surface with a second cation-containing solution, the cation being one or more of Cs.sup.+, Rb.sup.+, Ag.sup.+, Tl.sup.+, K.sup.+, and NH.sub.4.sup.+, thereby reducing amounts of radioactive cesium embedded in clays found on the surface.

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Billone, M. C.; Burtseva, T. A.

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).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lopes, Pedro; Vale, Alberto; Ventura, Rodrigo

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, themore » 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 Center of Rotation (ICR): a geometric approach where, at each time instant, the vehicle describes a circumference (either with a finite or infinite radius). The inverse of the kinematic model transforms the three input parameters of the center of mass into the four parameters for the wheels, preserving the omni-directional capabilities. The solution is implemented and tested using a HMI with a control disk and an analog joystick with two axis. The control disk was specially designed for this solution and implemented using a programmable micro-controller. In the first set of experiments, the HMI communicates with a computer running a simulator of the CPRHS/CTS, with the vehicle kinematics and dynamics, moving in a map of the ITER buildings. In the second set of experiments, the HMI communicates with a scaled prototype of the CPRHS running in a mock-up scenario to obtain more realistic results. Several type of tests were performed to evaluate the usability of the HMI. Different human operators without knowledge neither experience with this interface were invited to test the HMI. The operators had to drive the vehicle from an initial place to a final destination under the following conditions: with a pre-computed path to help guidance, without any path, with the information of the closest obstacles and without any help. The performance was evaluated using the time duration of the operation, the energy required to perform the described path, the risk of collision and, in case of a pre-computed path, the comparison between paths. In addition, each operator tested the HMI several times to evaluate the performance along consecutive trials. (authors)« less

Hybrid Skyshine Calculations for Complex Neutron and Gamma-Ray Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shultis, J. Kenneth

2000-10-15

A two-step hybrid method is described for computationally efficient estimation of neutron and gamma-ray skyshine doses far from a shielded source. First, the energy and angular dependence of radiation escaping into the atmosphere from a source containment is determined by a detailed transport model such as MCNP. Then, an effective point source with this energy and angular dependence is used in the integral line-beam method to transport the radiation through the atmosphere up to 2500 m from the source. An example spent-fuel storage cask is analyzed with this hybrid method and compared to detailed MCNP skyshine calculations.

Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Mo; Nakshatrala, Kalyana; William, Kasper

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels ofmore » protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Mo; Nakshatrala, Kalyana; William, Kasper

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels ofmore » protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.« less

KSC-2011-6651

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- The multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission, enclosed in a shipping cask, rolls into the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6658

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, a crane lifts the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission from its transportation pallet. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6647

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- The multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission, enclosed in a shipping cask, is seen through the open door of the MMRTG trailer that delivered it to the RTG storage facility at NASA's Kennedy Space Center in Florida. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6650

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- Workers use a forklift to transport the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission to the door of the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6648

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- Workers use a forklift to offload the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission from the MMRTG trailer that delivered it to the RTG storage facility at NASA's Kennedy Space Center in Florida. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6653

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, measurements are taken to determine the level of radioactivity emitted from the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission, enclosed in a shipping cask in the background. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6662

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, Department of Energy contractor employees remove the external and internal protective layers of the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6663

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, the external and internal protective layers of the shipping cask are lifted from around the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6649

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- Workers use a forklift to offload the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission from the MMRTG trailer that delivered it to the RTG storage facility at NASA's Kennedy Space Center in Florida. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6660

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission is lifted from around the MMRTG using guide rods installed on the support base. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

2013-07-01

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

Defense Remote Handled Transuranic Waste Cost/Schedule Optimization Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, G.D.; Beaulieu, D.H.; Wolaver, R.W.

1986-11-01

The purpose of this study is to provide the DOE information with which it can establish the most efficient program for the long management and disposal, in the Waste Isolation Pilot Plant (WIPP), of remote handled (RH) transuranic (TRU) waste. To fulfill this purpose, a comprehensive review of waste characteristics, existing and projected waste inventories, processing and transportation options, and WIPP requirements was made. Cost differences between waste management alternatives were analyzed and compared to an established baseline. The result of this study is an information package that DOE can use as the basis for policy decisions. As part ofmore » this study, a comprehensive list of alternatives for each element of the baseline was developed and reviewed with the sites. The principle conclusions of the study follow. A single processing facility for RH TRU waste is both necessary and sufficient. The RH TRU processing facility should be located at Oak Ridge National Laboratory (ORNL). Shielding of RH TRU to contact handled levels is not an economic alternative in general, but is an acceptable alternative for specific waste streams. Compaction is only cost effective at the ORNL processing facility, with a possible exception at Hanford for small compaction of paint cans of newly generated glovebox waste. It is more cost effective to ship certified waste to WIPP in 55-gal drums than in canisters, assuming a suitable drum cask becomes available. Some waste forms cannot be packaged in drums, a canister/shielded cask capability is also required. To achieve the desired disposal rate, the ORNL processing facility must be operational by 1996. Implementing the conclusions of this study can save approximately $110 million, compared to the baseline, in facility, transportation, and interim storage costs through the year 2013. 10 figs., 28 tabs.« less

WASTE HANDLING BUILDING ELECTRICAL SYSTEM DESCRIPTION DOCUMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.C. Khamamkar

2000-06-23

The Waste Handling Building Electrical System performs the function of receiving, distributing, transforming, monitoring, and controlling AC and DC power to all waste handling building electrical loads. The system distributes normal electrical power to support all loads that are within the Waste Handling Building (WHB). The system also generates and distributes emergency power to support designated emergency loads within the WHB within specified time limits. The system provides the capability to transfer between normal and emergency power. The system provides emergency power via independent and physically separated distribution feeds from the normal supply. The designated emergency electrical equipment will bemore » designed to operate during and after design basis events (DBEs). The system also provides lighting, grounding, and lightning protection for the Waste Handling Building. The system is located in the Waste Handling Building System. The system consists of a diesel generator, power distribution cables, transformers, switch gear, motor controllers, power panel boards, lighting panel boards, lighting equipment, lightning protection equipment, control cabling, and grounding system. Emergency power is generated with a diesel generator located in a QL-2 structure and connected to the QL-2 bus. The Waste Handling Building Electrical System distributes and controls primary power to acceptable industry standards, and with a dependability compatible with waste handling building reliability objectives for non-safety electrical loads. It also generates and distributes emergency power to the designated emergency loads. The Waste Handling Building Electrical System receives power from the Site Electrical Power System. The primary material handling power interfaces include the Carrier/Cask Handling System, Canister Transfer System, Assembly Transfer System, Waste Package Remediation System, and Disposal Container Handling Systems. The system interfaces with the MGR Operations Monitoring and Control System for supervisory monitoring and control signals. The system interfaces with all facility support loads such as heating, ventilation, and air conditioning, office, fire protection, monitoring and control, safeguards and security, and communications subsystems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bateman, V.I.; Brown, F.A.; Hansen, N.R.

Sandia National Laboratories (SNL) designs mechanical systems with electronics that must survive high shock environments. These mechanical systems include penetrators that must survive soil, rock, and ice penetration, nuclear transportation casks that must survive transportation environments, and laydown weapons that must survive delivery impact of 125 fps. These mechanical systems contain electronics that may operate during and after the high shock environment and that must be protected from the high shock environments. A study has been started to improve the packaging techniques for the advanced electronics utilized in these mechanical systems because current packaging techniques are inadequate for these moremore » sensitive electronics. In many cases, it has been found that the packaging techniques currently used not only do not mitigate the shock environment but actually amplify the shock environment. An ambitious goal for this packaging study is to avoid amplification and possibly attenuate the shock environment before it reaches the electronics contained in the various mechanical systems. As part of the investigation of packaging techniques, a two phase study of shock mitigating materials is being conducted. The purpose of the first phase reported here is to examine the performance of a joint that consists of shock mitigating material sandwiched in between steel and to compare the performance of the shock mitigating materials. A split Hopkinson bar experimental configuration simulates this joint and has been used to study the shock mitigating characteristics of seventeen, unconfined materials. The nominal input for these tests is an incident compressive wave with 50 fps peak (1,500 {micro}{var_epsilon} peak) amplitude and a 100 {micro}s duration (measured at 10% amplitude).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasparek, Eva M.; Voelzke, Holger; Scheidemann, Robert

Rigid, closed-cell polyurethane foams are frequently used as cask impact limiters in nuclear materials and hazardous waste transport due to their high energy-absorption potential. When assessing the cask integrity in accidental scenarios based on numerical simulations, a description of the foam damping properties is required for different strain rates and for a wide temperature range with respect to waste heat generation in conjunction with critical operating and environmental conditions. Implementation and adaption of a respective finite element material model strongly relies on an appropriate experimental data base. Even though extensive impact experiments were conducted e.g. in Sandia National Laboratories, Savannahmore » River National Laboratory and by Rolls Royce plc, not all relevant factors were taken into account. Hence, BAM who is in charge of the mechanical evaluation of such packages within the approval procedure in Germany, incorporated systematic test series into a comprehensive research project aimed to develop numerical methods for a couple of damping materials. In a first step, displacement driven compression tests have been performed on confined, cubic specimens at five loading rates ranging from 0.02 mm/s to 3 m/s at temperatures between +90 deg. C and -40 deg. C. Materials include two different polyurethane foam types called FR3718 and FR3730 having densities of 280 kg/m{sup 3} and 488 kg/m{sup 3} from the product line-up of General Plastics Manufacturing Company. Their data was used to adapt an advanced plasticity model allowing for reliably simulating cellular materials under multi-axial compression states. Therefore, an automated parameter identification procedure had been established by combining an artificial neural network with local optimization techniques. Currently, the selected numerical material input values are validated and optimized by means of more complex loading configurations with the prospect of establishing methods applicable to impact limiters under severe accidental conditions. The reference data base is provided by experiments, where weights between 212 kg and 1200 kg have been dropped from heights between 1.25 m and 7 m on confined 10 cm cubic foam specimens. By presenting the deviations between experimental values and the corresponding output of finite element simulations, the potentials and restrictions of the resulting models are highlighted. Systematic compression tests on polyurethane foams had been performed at BAM test site within the framework of a research project on impact limiters for handling casks for radioactive waste. The experimental results had been used to adapt numerical models for simulating the behaviour of different foam types at different temperatures. The loading speed, however, turned out to have a major influence on their flow curves that can not be captured by simple strain-rate dependent multipliers. Especially for guided drop tests that come close to real accidental scenarios there is a significant gap between experimental and numerical results even when applying such advanced material models. Hence, the extensive data base is currently deployed for expanding the standard algorithms to include adequate dynamic hardening factors. (authors)« less

Behavior of a tapered hub flange with a bolted flat cover in transient temperature field

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sawa, T.; Nakagomi, Y.; Hirose, T.

1996-02-01

When bolted flange connections with gaskets are used in mechanical structures such as pipe connections, bolted covers of casks, and pressure vessels in nuclear and chemical plants and cylinder heads in internal combustion engines, they are usually subjected to transient thermal conditions. An experimental and analytical study was made on a bolted connection subjected to thermal loading. The connection consists of an aluminum alloy tapered hub flange and a flat cover, including a gasket fastened by steel bolts and nuts. Temperature distribution in the connection was measured with thermocouples, and the axial bolt force, the maximum bolt stress, and themore » hub stress were measured by strain gages under a thermal condition that the inner surface of the flanges was heated and the outer surfaces of the flanges and the cover were held at room temperature. Finite difference analysis was made to obtain the temperature distributions in the connection due to a transient thermal condition. This paper demonstrates the method for obtaining an increment in axial bolt force and the maximum bolt stress. In all cases, the analytical results were fairly consistent with the experimental results.« less

Unique and massive Chernobyl cranes for deconstruction activities in the new safe confinement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parameswaran, N. A. Vijay; Chornyy, Igor; Owen, Rob

2013-07-01

On 26 April 1986, the worst nuclear power plant accident in history occurred at the Chernobyl plant in Ukraine (then part of the Soviet Union). The destruction of Unit 4 sent highly radioactive fallout over Belarus, Russia, Ukraine, and Europe. The object shelter-a containment sarcophagus-was built in November 1986 to limit exposure to radiation. However, it has only a planned 25-year lifespan and would probably not survive even a moderate seismic event in a region that has more than its share of such events. It was time to take action. One of the largest tasks that are in progress ismore » the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant Unit. One of the major mechanical handling systems to be installed in the new safe confinement is the Main Cranes System. The planned decontamination and decommissioning or dismantling activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the new safe confinement, will require large and sophisticated cranes. The article will focus on the current progress of the new safe confinement and of the main cranes system for the decommissioning or dismantling activities. (authors)« less

KSC-2011-6652

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- Workers reconnect the coolant hoses to the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission upon its arrival in the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida. Coolant flows through the hoses to dissipate any excess heat generated by the MMRTG. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6665

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, the external and internal protective layers of the shipping cask are lifted away from the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission. The MMRTG no longer needs supplemental cooling since any excess heat generated can dissipate into the air in the high bay. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6654

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, Innovative Health Applications employee Mike McPherson measures the level of radioactivity emitted from the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission, enclosed in a shipping cask at right. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6661

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, Innovative Health Applications employee David Lake measures the level of radioactivity emitted from the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission as the external protective layer of the shipping cask is removed. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

Probabilistic Multi-Hazard Assessment of Dry Cask Structures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bencturk, Bora; Padgett, Jamie; Uddin, Rizwan

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 environmentsmore » 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.« less

KSC-2011-6656

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, Department of Energy contractor employees attach cables to the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission during preparations to lift it from its transportation pallet. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6655

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, preparations are under way to attach the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission to the cables that will lift it from its transportation pallet. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

KSC-2011-6657

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, a Department of Energy contractor employee attaches a crane to the shipping cask enclosing the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission during preparations to lift it from its transportation pallet. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

The Storage, Transportation, and Disposal of Nuclear Waste

NASA Astrophysics Data System (ADS)

Younker, J. L.

2002-12-01

The U.S. Congress established a comprehensive federal policy to dispose of wastes from nuclear reactors and defense facilities, centered on deep geologic disposal of high-level radioactive waste. Site screening led to selection of three potential sites and in 1987, Congress directed the Secretary of Energy to characterize only one site: Yucca Mountain in Nevada. For more than 20 years, teams of scientists and engineers have been evaluating the potential suitability of the site. On the basis of their work, the U.S. Secretary of Energy, Spencer Abraham, concluded in February 2002 that a safe repository can be sited at Yucca Mountain. On July 23, 2002, President Bush signed Joint Resolution 87 approving the site at Yucca Mountain for development of a repository, which allows the U.S. Department of Energy (DOE) to prepare and submit a license application to the U.S. Nuclear Regulatory Commission (NRC). Concerns have been raised relative to the safe transportation of nuclear materials. The U.S. history of transportation of nuclear materials demonstrates that high-level nuclear materials can be safely transported. Since the 1960s, over 1.6 million miles have been traveled by more than 2,700 spent nuclear fuel shipments, and there has never been an accident severe enough to cause a release of radioactive materials. The DOE will use NRC-certified casks that must be able to withstand very stringent tests. The same design features that allow the casks to survive severe accidents also limit their vulnerability to sabotage. In addition, the NRC will approve all shipping routes and security plans. With regard to long-term safety, the Yucca Mountain disposal system has five key attributes. First, the arid climate and geology of Yucca Mountain combine to ensure that limited water will enter the emplacement tunnels. Second, the DOE has designed a waste package and drip shield that are expected to have very long lifetimes in the repository environment. Third, waste form solubilities limit radionuclide releases, and the invert material below the package would further delay radionuclide movement. Fourth, rock units in the unsaturated and saturated zone at Yucca Mountain will delay and dilute any radionuclides that have migrated away from the emplacement tunnels. Fifth, disruptions due to volcanism, seismic events, or nuclear criticality have been evaluated and all are shown to have very low likelihood of causing unacceptable doses. Volcanism could result in a small, but calculable, dose during the regulatory period of 10,000 years.

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feldkamp, Martin; Nehrig, Marko; Bletzer, Claus

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues ismore » provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)« less

Can Shale Safely Host U.S. Nuclear Waste?

NASA Astrophysics Data System (ADS)

Neuzil, C. E.

2013-07-01

Even as cleanup efforts after Japan's Fukushima disaster offer a stark reminder of the spent nuclear fuel (SNF) stored at nuclear plants worldwide, the decision in 2009 to scrap Yucca Mountain as a permanent disposal site has dimmed hope for a repository for SNF and other high-level nuclear waste (HLW) in the United States anytime soon. About 70,000 metric tons of SNF are now in pool or dry cask storage at 75 sites across the United States [Government Accountability Office, 2012], and uncertainty about its fate is hobbling future development of nuclear power, increasing costs for utilities, and creating a liability for American taxpayers [Blue Ribbon Commission on America's Nuclear Future, 2012].

ETF magnet design alternatives for the national MHD program

NASA Astrophysics Data System (ADS)

Marston, P. G.; Thome, R. J.; Dawson, A. M.; Bobrov, E. S.; Hatch, A. M.

1981-01-01

Five superconducting magnet designs are evaluated for a 200 MWe test facility requiring a magnet with an on-axis field of 6 T, an inlet bore area of 4 sq m, storing 6 x 10 to the 9th J. The designs include a straightforward rectangular saddle coil set, a 'Cask' configuration based on staves and corner blocks as the main support structure, and an internally cooled, cabled superconductor to minimize the substructure and eliminate the helium vessel. Also, a modular design using six coils with individual helium vessels and an integrated structure produces a simplest configuration which utilizes a natural rectangular interface for packaging the MHD channel and its connections, and results in a lower capital cost.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Barton L.

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 givenmore » the time and funds available« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bateman, V.I.; Bell, R.G. III; Brown, F.A.

Sandia National Laboratories (SNL) designs mechanical systems with electronics that must survive high shock environments. These mechanical systems include penetrators that must survive soil, rock, and ice penetration, nuclear transportation casks that must survive transportation environments, and laydown weapons that must survive delivery impact of 125-fps. These mechanical systems contain electronics that may operate during and after the high shock environment and that must be protected from the high shock environments. A study has been started to improve the packaging techniques for the advanced electronics utilized in these mechanical systems because current packaging techniques are inadequate for these more sensitivemore » electronics. In many cases, it has been found that the packaging techniques currently used not only do not mitigate the shock environment but actually amplify the shock environment. An ambitious goal for this packaging study is to avoid amplification and possibly attenuate the shock environment before it reaches the electronics contained in the various mechanical system. As part of the investigation of packaging techniques, a two part study of shock mitigating materials is being conducted. This paper reports the first part of the shock mitigating materials study. A study to compare three thicknesses (0.125, 0.250, and 0.500 in.) of seventeen, unconfined materials for their shock mitigating characteristics has been completed with a split Hopkinson bar configuration. The nominal input as measured by strain gages on the incident Hopkinson bar is 50 fps {at} 100 {micro}s for these tests. It is hypothesized that a shock mitigating material has four purposes: to lengthen the shock pulse, to attenuate the shock pulse, to mitigate high frequency content in the shock pulse, and to absorb energy. Both time domain and frequency domain analyses of the split Hopkinson bar data have been performed to compare the materials` achievement of these purposes.« less

The International Remote Monitoring Project: Results of the Swedish Nuclear Power Facility field trial

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, C.S.; af Ekenstam, G.; Sallstrom, M.

1995-07-01

The Swedish Nuclear Power Inspectorate (SKI) and the US Department of Energy (DOE) sponsored work on a Remote Monitoring System (RMS) that was installed in August 1994 at the Barseback Works north of Malmo, Sweden. The RMS was designed to test the front end detection concept that would be used for unattended remote monitoring activities. Front end detection reduces the number of video images recorded and provides additional sensor verification of facility operations. The function of any safeguards Containment and Surveillance (C/S) system is to collect information which primarily is images that verify the operations at a nuclear facility. Barsebackmore » is ideal to test the concept of front end detection since most activities of safeguards interest is movement of spent fuel which occurs once a year. The RMS at Barseback uses a network of nodes to collect data from microwave motion detectors placed to detect the entrance and exit of spent fuel casks through a hatch. A video system using digital compression collects digital images and stores them on a hard drive and a digital optical disk. Data and images from the storage area are remotely monitored via telephone from Stockholm, Sweden and Albuquerque, NM, USA. These remote monitoring stations operated by SKI and SNL respectively, can retrieve data and images from the RMS computer at the Barseback Facility. The data and images are encrypted before transmission. This paper presents details of the RMS and test results of this approach to front end detection of safeguard activities.« less

Performance of a personal neutron dosemeter based on direct ion storage at workplace fields in the nuclear industry.

PubMed

Boschung, M; Fiechtner, A; Wernli, C

2007-01-01

In the framework of the EVIDOS project, funded by the EC, measurements were carried out using dosemeters, based on ionisation chambers with direct ion storage (DIS-N), at several workplace fields, namely, at a fuel processing plant, a boiling and a pressurised water reactor, and near transport and storage casks. The measurements and results obtained with the DIS-N in these workplaces, which are representative for the nuclear industry, are described in this study. Different dosemeter configurations of converter and shielding materials were considered. The results are compared with values for personal dose equivalent which were assessed within the EVIDOS project by other partners. The advantages and limitations of the DIS-N dosemeter are discussed.

Industrial research for transmutation scenarios

NASA Astrophysics Data System (ADS)

Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

2011-04-01

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

The microbial diversity of an industrially produced lambic beer shares members of a traditionally produced one and reveals a core microbiota for lambic beer fermentation.

PubMed

Spitaels, Freek; Wieme, Anneleen D; Janssens, Maarten; Aerts, Maarten; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

2015-08-01

The microbiota involved in lambic beer fermentations in an industrial brewery in West-Flanders, Belgium, was determined through culture-dependent and culture-independent techniques. More than 1300 bacterial and yeast isolates from 13 samples collected during a one-year fermentation process were identified using matrix-assisted laser desorption/ionization time-of-flight mass spectrometry followed by sequence analysis of rRNA and various protein-encoding genes. The bacterial and yeast communities of the same samples were further analyzed using denaturing gradient gel electrophoresis of PCR-amplified V3 regions of the 16S rRNA genes and D1/D2 regions of the 26S rRNA genes, respectively. In contrast to traditional lambic beer fermentations, there was no Enterobacteriaceae phase and a larger variety of acetic acid bacteria were found in industrial lambic beer fermentations. Like in traditional lambic beer fermentations, Saccharomyces cerevisiae, Saccharomyces pastorianus, Dekkera bruxellensis and Pediococcus damnosus were the microorganisms responsible for the main fermentation and maturation phases. These microorganisms originated most probably from the wood of the casks and were considered as the core microbiota of lambic beer fermentations. Copyright © 2015 Elsevier Ltd. All rights reserved.

KSC-2011-6664

NASA Image and Video Library

2011-06-30

CAPE CANAVERAL, Fla. -- In the high bay of the RTG storage facility at NASA's Kennedy Space Center in Florida, a Department of Energy contractor employee guides the external and internal protective layers of the shipping cask as they are lifted from around the multi-mission radioisotope thermoelectric generator (MMRTG) for NASA's Mars Science Laboratory mission. The MMRTG no longer needs supplemental cooling since any excess heat generated can dissipate into the air in the high bay. The MMRTG will generate the power needed for the mission from the natural decay of plutonium-238, a non-weapons-grade form of the radioisotope. Heat given off by this natural decay will provide constant power through the day and night during all seasons. Waste heat from the MMRTG will be circulated throughout the rover system to keep instruments, computers, mechanical devices and communications systems within their operating temperature ranges. MSL's components include a compact car-sized rover, Curiosity, which has 10 science instruments designed to search for evidence on whether Mars has had environments favorable to microbial life, including chemical ingredients for life. The unique rover will use a laser to look inside rocks and release its gasses so that the rover’s spectrometer can analyze and send the data back to Earth. Launch of MSL aboard a United Launch Alliance Atlas V rocket is scheduled for Nov. 25 from Space Launch Complex 41 on Cape Canaveral Air Force Station in Florida. For more information, visit http://www.nasa.gov/msl. Photo credit: NASA/Frankie Martin

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matloch, L.; Vaccaro, S.; Couland, M.

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

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

2013-07-01

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)« less

SNF Interim Storage Canister Corrosion and Surface Environment Investigations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryan, Charles R.; Enos, David G.

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 presentmore » through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.« less

Advanced Borobond™ Shields for Nuclear Materials Containment and Borobond™ Immobilization of Volatile Fission Products - Final CRADA Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagh, Arun S.

2016-05-19

Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such asmore » Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.« less

The Microbial Diversity of Traditional Spontaneously Fermented Lambic Beer

PubMed Central

Spitaels, Freek; Wieme, Anneleen D.; Janssens, Maarten; Aerts, Maarten; Daniel, Heide-Marie; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

2014-01-01

Lambic sour beers are the products of a spontaneous fermentation that lasts for one to three years before bottling. The present study determined the microbiota involved in the fermentation of lambic beers by sampling two fermentation batches during two years in the most traditional lambic brewery of Belgium, using culture-dependent and culture-independent methods. From 14 samples per fermentation, over 2000 bacterial and yeast isolates were obtained and identified. Although minor variations in the microbiota between casks and batches and a considerable species diversity were found, a characteristic microbial succession was identified. This succession started with a dominance of Enterobacteriaceae in the first month, which were replaced at 2 months by Pediococcus damnosus and Saccharomyces spp., the latter being replaced by Dekkera bruxellensis at 6 months fermentation duration. PMID:24748344

The microbial diversity of traditional spontaneously fermented lambic beer.

PubMed

Spitaels, Freek; Wieme, Anneleen D; Janssens, Maarten; Aerts, Maarten; Daniel, Heide-Marie; Van Landschoot, Anita; De Vuyst, Luc; Vandamme, Peter

2014-01-01

Lambic sour beers are the products of a spontaneous fermentation that lasts for one to three years before bottling. The present study determined the microbiota involved in the fermentation of lambic beers by sampling two fermentation batches during two years in the most traditional lambic brewery of Belgium, using culture-dependent and culture-independent methods. From 14 samples per fermentation, over 2000 bacterial and yeast isolates were obtained and identified. Although minor variations in the microbiota between casks and batches and a considerable species diversity were found, a characteristic microbial succession was identified. This succession started with a dominance of Enterobacteriaceae in the first month, which were replaced at 2 months by Pediococcus damnosus and Saccharomyces spp., the latter being replaced by Dekkera bruxellensis at 6 months fermentation duration.

FINAL SAFETY ANALYSIS REPORT--SNAP 1A RADIOISOTOPE FUELED THERMOELECTRIC GENERATOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.P.

1960-06-30

The safety aspects involved in utilizing the Task 2 radioisotope-powered thermoelectric generator in a terrestrial satellite are described. It is based upon a generalized satellite mission having a 600-day orbital lifetime. A description of the basic design of the generator is presented in order to establish the analytical model. This includes the generator design, radiocerium fuel properties, and the fuel core. The transport of the generator to the launch site is examined, including the shipping cask, shipping procedures, and shipping hazards. A description of ground handling and vehicle integration is presented including preparation for fuel transfer, transfer, mating of generatorsmore » to final stage, mating final stage to booster, and auxiliary support equipment. The flight vehicle is presented to complete the analytical model. Contained in this chapter are descriptions of the booster-sustainer, final stage, propellants, and built-in safety systems. The typical missile range is examined with respect to the launch complex and range safety characteristics. The shielding of the fuel is discussed and includes both dose rates and shield thicknesses required. The bare core, shielded generator, fuel transfer operation and dose rates for accidental conditions are treated. mechanism of re-entry from the successful mission is covered. Radiocerium inventories with respect to time and the chronology of re-entry are specifically treated. The multiplicity of conditions for aborted missions is set forth. The definition of aborted missions is treated first in order to present the initial conditions. Following this, a definition of the forces imposed upon the generator is presented. The aborted missions is presented. A large number of initial vehicle failure cases is narrowed down into categories of consequences. Since stratospheric injection of fuel results in cases where the fuel is not contained after re-entry, an extensive discussion of the fall-out mechanism is presented. (auth)« less

Nondestructive Examination Guidance for Dry Storage Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meyer, Ryan M.; Suffield, Sarah R.; Hirt, Evelyn H.

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 ofmore » the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure--Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific components.« less

Type B drum packages

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCoy, J.C.

1994-08-01

The Type B drum packages (TBD) are conceptualized as a family of containers in which a single 208 L or 114 L (55 gal or 30 gal) drum containing Type B quantities of radioactive material (RAM) can be packaged for shipment. The TBD containers are being developed to fill a void in the packaging and transportation capabilities of the U.S. Department of Energy as no container packaging single drums of Type B RAM exists offering double containment. Several multiple-drum containers currently exist, as well as a number of shielded casks, but the size and weight of these containers present manymore » operational challenges for single-drum shipments. As an alternative, the TBD containers will offer up to three shielded versions (light, medium, and heavy) and one unshielded version, each offering single or optional double containment for a single drum. To reduce operational complexity, all versions will share similar design and operational features where possible. The primary users of the TBD containers are envisioned to be any organization desiring to ship single drums of Type B RAM, such as laboratories, waste retrieval activities, emergency response teams, etc. Currently, the TBD conceptual design is being developed with the final design and analysis to be completed in 1995 to 1996. Testing and certification of the unshielded version are planned to be completed in 1996 to 1997 with production to begin in 1997 to 1998.« less

Proceedings of the Chornobyl phytoremediation and biomass energy conversion workshop (in English;Russian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hartley, J.; Tokarevsky, V.

1998-06-01

Many concepts, systems, technical approaches, technologies, ideas, agreements, and disagreements were vigorously discussed during the course of the 2-day workshop. The workshop was successful in generating intensive discussions on the merits of the proposed concept that includes removal of radionuclides by plants and trees (phytoremediation) to clean up soil in the Chornobyl Exclusion Zone (CEZ), use of the resultant biomass (plants and trees) to generate electrical power, and incorporation of ash in concrete casks to be used as storage containers in a licensed repository for low-level waste. Twelve years after the Chornobyl Nuclear Power Plant (ChNPP) Unit 4 accident, whichmore » occurred on April 26, 1986, the primary 4radioactive contamination of concern is from radioactive cesium ({sup 137}Cs) and strontium ({sup 90}Sr). The {sup 137}Cs and {sup 90}Sr were widely distributed throughout the CEZ. The attendees from Ukraine, Russia, Belarus, Denmark and the US provided information, discussed and debated the following issues considerably: distribution and characteristics of radionuclides in CEZ; efficacy of using trees and plants to extract radioactive cesium (Cs) and strontium (Sr) from contaminated soil; selection of energy conversion systems and technologies; necessary infrastructure for biomass harvesting, handling, transportation, and energy conversion; radioactive ash and emission management; occupational health and safety concerns for the personnel involved in this work; and economics. The attendees concluded that the overall concept has technical and possibly economic merits. However, many issues (technical, economic, risk) remain to be resolved before a viable commercial-scale implementation could take place.« less

Remote Whispering Applying Time Reversal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, Brian Eric

The purpose of this project was to explore the use of time reversal technologies as a means for communication to a targeted individual or location. The idea is to have the privacy of whispering in one’s ear, but to do this remotely from loudspeakers not located near the target. Applications of this work include communicating with hostages and survivors in rescue operations, communicating imaging and operational conditions in deep drilling operations, monitoring storage of spent nuclear fuel in storage casks without wires, or clandestine activities requiring signaling between specific points. This technology provides a solution in any application where wiresmore » and radio communications are not possible or not desired. It also may be configured to self calibrate on a regular basis to adjust for changing conditions. These communications allow two people to converse with one another in real time, converse in an inaudible frequency range or medium (i.e. using ultrasonic frequencies and/or sending vibrations through a structure), or send information for a system to interpret (even allowing remote control of a system using sound). The time reversal process allows one to focus energy to a specific location in space and to send a clean transmission of a selected signal only to that location. In order for the time reversal process to work, a calibration signal must be obtained. This signal may be obtained experimentally using an impulsive sound, a known chirp signal, or other known signals. It may also be determined from a numerical model of a known environment in which the focusing is desired or from passive listening over time to ambient noise.« less

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less

Radioactive waste material melter apparatus

DOEpatents

Newman, D.F.; Ross, W.A.

1990-04-24

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J

2016-01-01

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blademore » histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.« less

Technology, safety and costs of decommissioning reference independent spent fuel storage installations. [Contains glossary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludwick, J D; Moore, E B

1984-01-01

Safety and cost information is developed for the conceptual decommissioning of five different types of reference independent spent fuel storage installations (ISFSIs), each of which is being given consideration for interim storage of spent nuclear fuel in the United States. These include one water basin-type ISFSI (wet) and four dry ISFSIs (drywell, silo, vault, and cask). The reference ISFSIs include all component parts necessary for the receipt, handling and storage of spent fuel in a safe and efficient manner. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, and potential radiation doses tomore » the public. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment followed by long-term surveillance).« less

Apparatus for safeguarding a radiological source

DOEpatents

Bzorgi, Fariborz M

2014-10-07

A tamper detector is provided for safeguarding a radiological source that is moved into and out of a storage location through an access porthole for storage and use. The radiological source is presumed to have an associated shipping container approved by the U.S. Nuclear Regulatory Commission for transporting the radiological source. The tamper detector typically includes a network of sealed tubing that spans at least a portion of the access porthole. There is an opening in the network of sealed tubing that is large enough for passage therethrough of the radiological source and small enough to prevent passage therethrough of the associated shipping cask. Generally a gas source connector is provided for establishing a gas pressure in the network of sealed tubing, and a pressure drop sensor is provided for detecting a drop in the gas pressure below a preset value.

Radioactive waste material melter apparatus

DOEpatents

Newman, Darrell F.; Ross, Wayne A.

1990-01-01

An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

The Direct Path To WIPP - 12471

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spoerner, M.T.; Burger, M.J.; Garcia, J.

2012-07-01

Sandia National Laboratories/New Mexico (SNL/NM), designated as a small quantity site (SQS) by the National TRU Program (NTP), generated contact-handled (CH) and remote-handled (RH) transuranic (TRU) waste primarily from the decontamination and clean-out of glove boxes at the Hot Cell Facility (HCF) at Technical Area (TA) V. All of the waste required repackaging, with the CH TRU waste being repackaged from late 2007 through 2011. Three shipments of CH were completed in October 2011, which de-inventoried SNL/NM's legacy TRU waste. In FY11, RH TRU waste was repackaged at the Auxiliary Hot Cell Facility (AHCF) located in TAV with the supportmore » of the Central Characterization Project (CCP). The waste was originally packaged in SNL/NM fabricated casks, cement or lead-lined 55-gallon drums, or 30-gallon drums. The AHCF is a small hot cell, with access only through a roof port which presented challenges for inserting and removing waste from the hot cell. The CCP provided visual examination operators (VEOs) to observe and document each waste item repackaged, removal of prohibited items, and radiological sampling. Dose-to-Curie measurements were calculated by CCP after a radiological report was prepared using scaling factors determined by the analysis of swipe samples. Finally, headspace gas samples were taken and sent to the Advanced Mixed Waste Treatment Project (AMWTP) for analysis. Despite the challenges, the RH waste is on track to be shipped to WIPP in early FY12. The processes used and procedures developed to conduct the repackaging operations, the issues identified and mitigated were challenging but the cooperation between SNL/NM and the Central Characterization Program (CCP) enabled SNL/NM to complete the repackaging and support the characterization and shipment. An inventory list, identification of the campaigns, discussion of the challenges and mitigations, and the final loading of the RH 72-B casks at TA-V for direct shipment to the Waste Isolation Pilot Plant (WIPP) will be discussed. Lessons learned from the RH campaigns are: - Some containers that were originally identified as HC-3 have been re-evaluated and became < HC-3 due to the conservative estimates made by the original generators - Operators at the AHCF were not accustomed to the detail required by the VE operators. However, they worked well together and the repackaging was completed ahead of schedule. - The AK was not always accurate as was demonstrated by the solid waste found in the drum during the first visit by EPA. That waste has since been determined to be low-level. - Two drums originally thought to be RH turned out to be CH and arrangement for RTR had to be made quickly. - Six of the original RH repacked drums became low level. - Lessons learned from the CH campaigns were helpful in avoiding many issues. The RH repackaging effort has been a success due to the expertise of the AHCF operators, supervisor, and manager, the conscientious attention to detail of the CCP VE operators, the experience of the CCP DTC and headspace gas sampling staff, and the guidance and support from CCP and CBFO. Sometimes schedules had to be adjusted, processes updated, and issues discussed, but the communication between CCP and SNL/NM was good. SNL/NM hopes to have the legacy RH TRU waste shipped off-site by early 2012. (authors)« less

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

Price elasticity of on- and off-premises demand for alcoholic drinks: A Tobit analysis.

PubMed

Jiang, Heng; Livingston, Michael; Room, Robin; Callinan, Sarah

2016-06-01

Understanding how price policies will affect alcohol consumption requires estimates of the impact of price on consumption among different types of drinkers and across different consumption settings. This study aims to estimate how changes in price could affect alcohol demand across different beverages, different settings (on-premise, e.g., bars, restaurants and off-premise, e.g., liquor stores, supermarkets), and different levels of drinking and income. Tobit analysis is employed to estimate own- and cross-price elasticities of alcohol demand among 11 subcategories of beverage based on beverage type and on- or off-premise supply, using cross-sectional data from the Australian arm of the International Alcohol Control Survey 2013. Further elasticity estimates were derived for sub-groups of drinkers based on their drinking and income levels. The results suggest that demand for nearly every subcategory of alcohol significantly responds to its own price change, except for on-premise spirits and ready-to-drink spirits. The estimated demand for off-premise beverages is more strongly affected by own price changes than the same beverages in on-premise settings. Demand for off-premise regular beer and off-premise cask wine is more price responsive than demand for other beverages. Harmful drinkers and lower income groups appear more price responsive than moderate drinkers and higher income groups. Our findings suggest that alcohol price policies, such as increasing alcohol taxes or introducing a minimum unit price, can reduce alcohol demand. Price appears to be particularly effective for reducing consumption and as well as alcohol-related harm among harmful drinkers and lower income drinkers. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.

This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportationmore » of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an 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 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF 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 licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.« less

Three dimensional contact/impact methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

Remanent Activation in the Mini-SHINE Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Micklich, Bradley J.

2015-04-16

Argonne National Laboratory is assisting SHINE Medical Technologies in developing a domestic source of the medical isotope 99Mo through the fission of low-enrichment uranium in a uranyl sulfate solution. In Phase 2 of these experiments, electrons from a linear accelerator create neutrons by interacting in a depleted uranium target, and these neutrons are used to irradiate the solution. The resulting neutron and photon radiation activates the target, the solution vessels, and a shielded cell that surrounds the experimental apparatus. When the experimental campaign is complete, the target must be removed into a shielding cask, and the experimental components must bemore » disassembled. The radiation transport code MCNPX and the transmutation code CINDER were used to calculate the radionuclide inventories of the solution, the target assembly, and the shielded cell, and to determine the dose rates and shielding requirements for selected removal scenarios for the target assembly and the solution vessels.« less

Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

NASA Image and Video Library

1969-11-19

AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

Performance of the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities.

PubMed

Lahaye, T; Chau, Q; Ménard, S; Lacoste, V; Muller, H; Luszik-Bhadra, M; Reginatto, M; Bruguier, P

2006-01-01

This paper mainly aims at presenting the measurements and the results obtained with the electronic personal neutron dosemeter Saphydose-N at different facilities. Three campaigns were led in the frame of the European contract EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The first one consisted in the measurements at the IRSN French research laboratory in reference neutron fields generated by a thermal facility (SIGMA), radionuclide ISO sources ((241)AmBe; (252)Cf; (252)Cf(D(2)O)\\Cd) and a realistic spectrum (CANEL/T400). The second one was performed at the Krümmel Nuclear Power Plant (Germany) close to the boiling water reactor and to a spent fuel transport cask. The third one was realised at Mol (Belgium), at the VENUS Research Reactor and at Belgonucléaire, a fuel processing factory.

Assessment of Impact of Monoenergetic Photon Sources on Prioritized Nonproliferation Applications: Simulation Study Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geddes, Cameron; Ludewigt, Bernhard; Valentine, John

Near-monoenergetic photon sources (MPSs) have the potential to improve sensitivity at greatly reduced dose in existing applications and enable new capabilities in other applications. MPS advantages include the ability to select energy, energy spread, flux, and pulse structures to deliver only the photons needed for the application, while suppressing extraneous dose and background. Some MPSs also offer narrow divergence photon beams which can target dose and/or mitigate scattering contributions to image contrast degradation. Current broad-band, bremsstrahlung photon sources (e.g., linacs and betatrons) deliver unnecessary dose that in some cases also interferes with the signature to be detected and/or restricts operations,more » and must be collimated (reducing flux) to generate narrow divergence beams. While MPSs can in principle resolve these issues, they are technically challenging to produce. Candidate MPS technologies for nonproliferation applications are now being developed, each of which have different properties (e.g. broad divergence vs. narrow). Within each technology, source parameters trade off against one another (e.g. flux vs. energy spread), representing a large operation space. To guide development, requirements for each application of interest must be defined and simulations conducted to define MPS parameters that deliver benefit relative to current systems. The present project conducted a broad assessment of potential nonproliferation applications where MPSs may provide new capabilities or significant performance enhancement (reported separately), which led to prioritization of several applications for detailed analysis. The applications prioritized were: cargo screening and interdiction of Special Nuclear Materials (SNM), detection of hidden SNM, treaty/dismantlement verification, and spent fuel dry storage cask content verification. High resolution imaging for stockpile stewardship was considered as a sub-area of the treaty topic, as it is also of interest for future treaty use. This report presents higher-fidelity calculations and modeling results to quantitatively evaluate the prioritized applications, and to derive the key MPS properties that drive application benefit. Simulations focused on the conventional signatures of radiography, photofission, and NRF to enable comparison to present methods and evaluation of benefit.« less

Deterministically estimated fission source distributions for Monte Carlo k-eigenvalue problems

DOE PAGES

Biondo, Elliott D.; Davidson, Gregory G.; Pandya, Tara M.; ...

2018-04-30

The standard Monte Carlo (MC) k-eigenvalue algorithm involves iteratively converging the fission source distribution using a series of potentially time-consuming inactive cycles before quantities of interest can be tallied. One strategy for reducing the computational time requirements of these inactive cycles is the Sourcerer method, in which a deterministic eigenvalue calculation is performed to obtain an improved initial guess for the fission source distribution. This method has been implemented in the Exnihilo software suite within SCALE using the SPNSPN or SNSN solvers in Denovo and the Shift MC code. The efficacy of this method is assessed with different Denovo solutionmore » parameters for a series of typical k-eigenvalue problems including small criticality benchmarks, full-core reactors, and a fuel cask. Here it is found that, in most cases, when a large number of histories per cycle are required to obtain a detailed flux distribution, the Sourcerer method can be used to reduce the computational time requirements of the inactive cycles.« less

How we shipped our flip and standard too

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deigl, H.J.; Feltz, D.E.

1984-07-01

This paper highlights the planning and handling activities for the shipment of irradiated TRIGA fuel from Texas A and M University to the Argonne National Lab/West (ANL/West) reactor facility at Idaho Falls, Idaho. Attention is focused on the enormous time spent on the planning and preparations prior to the shipment. The actual handling time at the NSCR for three shipping packages containing a total 51 elements was only 4 days, but, the time spent in planning and preparation exceeded 16 months. The fuel was transferred for shipment without incident - and from a health physics standpoint the exercise went verymore » well. Whole body exposures and hand doses were minimal for such a large undertaking. ANL/West health physicists reported contamination of the lifting devices for the HFIR when they received the cask. These pieces were wipe tested and contamination was found to be less than 200 dpm. If they were contaminated we were extremely fortunate during handling not to contaminate our facility or personnel.« less

Antineutrino Monitoring of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Brdar, Vedran; Huber, Patrick; Kopp, Joachim

2017-11-01

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 worldwide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this paper, 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 reverify 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 contaminated areas such as the Hanford site in Washington state.

Preparation for Testing, Safe Packing and Shipping of Spent Nuclear Fuel from IFIN-HH, Bucharest-Magurele to Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dragolici, C.A.; Zorliu, A.; Popa, V.

2007-07-01

The Russian Research Reactor Fuel Return (RRRFR) program is promoted by IAEA and DOE in order to repatriate of irradiated research reactor fuel originally supplied by Russia to facilities outside the country. Developed under the framework of the Global Threat Reduction Initiative (GTRI) the take-back program [1] common goal is to reduce both proliferation and security risks by eliminating or consolidating inventories of high-risk material. The main objective of this program is to support the return to Russian Federation of fresh or irradiated HEU and LEU fuel. Being part of this project, Romania is fulfilling its tasks by examining transportmore » and transfer cask options, assessment of transport routes, and providing cost estimates for required equipment and facility modifications. Spent Nuclear Fuel (SNF) testing, handling, packing and shipping are the most common interests on which the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei' (IFIN-HH) is focusing at the moment. (authors)« less

Deterministically estimated fission source distributions for Monte Carlo k-eigenvalue problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Biondo, Elliott D.; Davidson, Gregory G.; Pandya, Tara M.

The standard Monte Carlo (MC) k-eigenvalue algorithm involves iteratively converging the fission source distribution using a series of potentially time-consuming inactive cycles before quantities of interest can be tallied. One strategy for reducing the computational time requirements of these inactive cycles is the Sourcerer method, in which a deterministic eigenvalue calculation is performed to obtain an improved initial guess for the fission source distribution. This method has been implemented in the Exnihilo software suite within SCALE using the SPNSPN or SNSN solvers in Denovo and the Shift MC code. The efficacy of this method is assessed with different Denovo solutionmore » parameters for a series of typical k-eigenvalue problems including small criticality benchmarks, full-core reactors, and a fuel cask. Here it is found that, in most cases, when a large number of histories per cycle are required to obtain a detailed flux distribution, the Sourcerer method can be used to reduce the computational time requirements of the inactive cycles.« less

FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schindelholz, Eric John; Bryan, Charles R.; Alexander, Christopher L.

This progress report describes work done in FY17 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. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosionmore » work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.« less

Exome Pool-Seq in neurodevelopmental disorders.

PubMed

Popp, Bernt; Ekici, Arif B; Thiel, Christian T; Hoyer, Juliane; Wiesener, Antje; Kraus, Cornelia; Reis, André; Zweier, Christiane

2017-12-01

High throughput sequencing has greatly advanced disease gene identification, especially in heterogeneous entities. Despite falling costs this is still an expensive and laborious technique, particularly when studying large cohorts. To address this problem we applied Exome Pool-Seq as an economic and fast screening technology in neurodevelopmental disorders (NDDs). Sequencing of 96 individuals can be performed in eight pools of 12 samples on less than one Illumina sequencer lane. In a pilot study with 96 cases we identified 27 variants, likely or possibly affecting function. Twenty five of these were identified in 923 established NDD genes (based on SysID database, status November 2016) (ACTB, AHDC1, ANKRD11, ATP6V1B2, ATRX, CASK, CHD8, GNAS, IFIH1, KCNQ2, KMT2A, KRAS, MAOA, MED12, MED13L, RIT1, SETD5, SIN3A, TCF4, TRAPPC11, TUBA1A, WAC, ZBTB18, ZMYND11), two in 543 (SysID) candidate genes (ZNF292, BPTF), and additionally a de novo loss-of-function variant in LRRC7, not previously implicated in NDDs. Most of them were confirmed to be de novo, but we also identified X-linked or autosomal-dominantly or autosomal-recessively inherited variants. With a detection rate of 28%, Exome Pool-Seq achieves comparable results to individual exome analyses but reduces costs by >85%. Compared with other large scale approaches using Molecular Inversion Probes (MIP) or gene panels, it allows flexible re-analysis of data. Exome Pool-Seq is thus well suited for large-scale, cost-efficient and flexible screening in characterized but heterogeneous entities like NDDs.

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.O. Bader

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are releasedmore » from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.« less

Romania: Brand-New Engineering Solutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ken Allen; Lucian Biro; Nicolae Zamfir

The HEU spent nuclear fuel transport from Romania was a pilot project in the framework of the Russian Research Reactor Fuel Return Program (RRRFR), being the first fully certified spent nuclear fuel shipment by air. The successful implementation of the Romanian shipment also brought various new technology in the program, further used by other participating countries. Until 2009, the RRRFR program repatriated to the Russian Federation HEU spent nuclear fuel of Russian origin from many countries, like Uzbekistan, Czech Republic, Latvia, Hungary, Kazakhstan and Bulgaria. The means of transport used were various; from specialized TK-5 train for the carriage ofmore » Russian TUK-19 transport casks, to platform trains for 20 ft freight ISO containers carrying Czech Skoda VPVR/M casks; from river barge on the Danube, to vessel on the Mediterranean Sea and Atlantic Ocean. Initially, in 2005, the transport plan of the HEU spent nuclear fuel from the National Institute for R&D in Nuclear Physics and Nuclear Engineering 'Horia Hulubei' in Magurele, Romania considered a similar scheme, using the specialized TK-5 train transiting Ukraine to the destination point in the Russian Federation, or, as an alternative, using the means and route of the spent nuclear fuel periodically shipped from the Bulgarian nuclear power plant Kosloduy (by barge on the Danube, and by train through Ukraine to the Russian Federation). Due to impossibility to reach an agreement in due time with the transit country, in February 2007 the US, Russian and Romanian project partners decided to adopt the air shipment of the spent nuclear fuel as prime option, eliminating the need for agreements with any transit countries. By this time the spent nuclear fuel inspections were completed, proving the compliance of the burn-up parameters with the international requirements for air shipments of radioactive materials. The short air route avoiding overflying of any other countries except the country of origin and the country of destination also contributed to the decision making in this issue. The efficient project management and cooperation between the three countries (Russia, Romania and USA) made possible, after two and a half years of preparation work, for the first fully certified spent nuclear fuel air shipment to take place on 29th of June 2009, from Romanian airport 'Henri Coanda' to the Russian airport 'Koltsovo' near Yekaterinburg. One day before that, after a record period of 3 weeks of preparation, another HEU cargo was shipped by air from Romanian Institute for Nuclear Research in Pitesti to Russia, containing fresh pellets and therefore making Romania the third HEU-free country in the RRRFR program.« less

The effect of stress state on zirconium hydride reorientation

NASA Astrophysics Data System (ADS)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by correlating the finite element stress-state results with the spatial distribution of hydride microstructures observed within the optical micrographs for each sample. Experiments showed that the hydride reorientation was enhanced as the stress biaxiality increased. The threshold stress decreased from 150 MPa to 80 MPa when stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension. This behavior was also predicted by classical nucleation theory based on the Gibbs free energy of transformation being assisted by the far-field stress. An analysis of in situ X-ray diffraction data obtained during a thermo-mechanical cycle typical of vacuum drying showed a complex lattice-spacing behavior of the hydride phase during the dissolution and precipitation. The in-plane hydrides showed bilinear lattice expansion during heating with the intrinsic thermal expansion rate of the hydrides being observed only at elevated temperatures as they dissolve. For radial hydrides that precipitate during cooling under stress, the spacing of the close-packed {111} planes oriented normal to the maximum applied stress was permanently higher than the corresponding {111} plane spacing in the other directions. This behavior is believed to be a result of a complex stress state within the precipitating plate-like hydrides that induces a strain component within the hydrides normal to its "plate" face (i.e., the applied stress direction) that exceeds the lattice spacing strains in the other directions. During heat-up, the lattice spacing of these same "plate" planes actually contract due to the reversion of the stress state within the plate-like hydrides as they dissolve. The presence of radial hydrides and their connectivity with in-plane hydrides was shown to increase the ductile-to-brittle transition temperature during tensile testing. This behavior can be understood in terms of the role of radial hydrides in promoting the initiation of a long crack that subsequently propagates under fracture mechanics conditions. Finally, the d-spacing of irradiated Zircaloy-4 and M5 cladding tubes was measured at room temperature and compared to that of unirradiated samples.

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Off-premise alcohol purchasing in Australia: Variations by age group, income level and annual amount purchased.

PubMed

Jiang, Heng; Callinan, Sarah; Livingston, Michael; Room, Robin

2017-03-01

To delineate what type and how much alcohol is purchased from different types of off-licence premises and how this varies across demographic sub-groups, as a basis for public debate and decisions on pricing and planning policies to reduce alcohol-related harm in Australia. The data on alcohol purchasing from off-licence premises are taken from the Australian Alcohol Consumption and Purchasing survey-a nationally representative landline and mobile telephone survey in 2013 on the experiences with alcohol consumption and purchasing of 2020 Australians aged 16+. The present analysis uses data from 1730 respondents who purchased alcohol from off-licence premises in the previous 6 months. The majority (54%) of alcohol purchased from off-licence premises was sold from liquor barns (large warehouse-style alcohol stores), with bottle shops (31%) the second most common outlet. Cask wine was the cheapest alcohol available at off-licence premises in Australia. Respondents in higher alcohol purchasing quintiles and with those with lower income purchased a higher percentage of cheaper alcohol in their total volume of purchasing than lower purchasing quintiles and those with middle and higher income, and younger respondents purchased more expensive alcohol than older age groups. A minimum unit price or increasing alcohol taxes may effectively reduce alcohol purchasing for lower income heavy alcohol purchasers and older age groups from off-licence premise sources, and may be less effective on younger age groups. [Jiang H, Callinan S, Livingston M, Room R. Off-premise alcohol purchasing in Australia: Variations by age group, income level and annual amount purchased. Drug Alcohol Rev 2017;36:210-219]. © 2016 Australasian Professional Society on Alcohol and other Drugs.

Neutron detection devices with 6LiF converter layers

NASA Astrophysics Data System (ADS)

Finocchiaro, Paolo; Cosentino, Luigi; Meo, Sergio Lo; Nolte, Ralf; Radeck, Desiree

2018-01-01

The demand for new thermal neutron detectors as an alternative to 3He tubes in research, industrial, safety and homeland security applications, is growing. These needs have triggered research and development activities about new generations of thermal neutron detectors, characterized by reasonable efficiency and gamma rejection comparable to 3He tubes. In this paper we show the state of art of a promising lowcost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. Several configurations were studied with the GEANT4 simulation code, and then calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured detection efficiency is well reproduced by the simulations, therefore validating the simulation tool in view of new designs. These neutron detectors have also been tested at neutron beam facilities like ISIS (Rutherford Appleton Laboratory, UK) and n_TOF (CERN) where a few samples are already in operation for beam flux and 2D profile measurements. Forthcoming applications are foreseen for the online monitoring of spent nuclear fuel casks in interim storage sites.

IDAHO NATIONAL LABORATORY TRANSPORTATION TASK REPORT ON ACHIEVING MODERATOR EXCLUSION AND SUPPORTING STANDARDIZED TRANSPORTATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2011-09-01

Following the defunding of the Yucca Mountain Project, it is reasonable to assume that commercial used fuel will remain in storage for the foreseeable future. This report proposes supplementing the ongoing research and development work related to potential degradation of used fuel, baskets, poisons, and storage canisters during an extended period of storage with a parallel path. This parallel path can assure criticality safety during transportation by implementing a concept that achieves moderator exclusion (no in-leakage of moderator into the used fuel cavity). Using updated risk assessment insights for additional technical justification and relying upon a component inside of themore » transportation cask that provides a watertight function, a strong argument can be made that moderator intrusion is not credible and should not be a required assumption for criticality evaluations during normal conditions of transportation. A demonstrating testing program supporting a detailed analytical effort as well as updated risk assessment insights can provide the basis for moderator exclusion during hypothetical accident conditions. This report also discusses how this engineered concept can support the goal of standardized transportation.« less

Polysaccharides and lignin from oak wood used in cooperage: Composition, interest, assays: A review.

PubMed

Le Floch, Alexandra; Jourdes, Michael; Teissedre, Pierre-Louis

2015-11-19

It is widely accepted that alcoholic beverage quality depends on their ageing in premium quality oak wood. From the choice of wood to beverage ageing, through the different steps in cask manufacturing, many factors should be considered. One of the biggest challenge in cooperages is to take into account all these factors. Most of the studies are interested in phenolic compounds, extracted during ageing and especially involved in wine oxidation, colour, and sensory properties such as astringency and bitterness. Oak aroma volatile compounds have also been the subject of numerous studies. These compounds of interest are part of low molecular weight compounds which represent 2%-10% of oak wood composition. However, three polymers constitute the main part of oak wood: cellulose, hemicellulose and lignin. As far as we are aware, few studies concerning the role of these major macromolecules in oak wood have been published previously. This article reviews oak wood polysaccharides and lignin, their potential interest and different assays used to determine their content. Copyright © 2015 Elsevier Ltd. All rights reserved.

Prototype pushing robot for emplacing vitrified waste canisters into horizontal disposal drifts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Londe, L.; Seidler, W.K.; Bosgiraud, J.M.

2007-07-01

Within the French Underground Disposal concept, as described in ANDRA's (Agence Nationale pour la Gestion des Dechets Radioactifs) Dossier 2005, the Pushing Robot is an application envisaged for the emplacement (and the potential retrieval) of 'Vitrified waste packages', also called 'C type packages'. ANDRA has developed a Prototype Pushing Robot within the framework of the ESDRED Project (Engineering Studies and Demonstration of Repository Design) which is co-funded by the European Commission as part of the sixth EURATOM Research and Training Framework Programme (FP6) on nuclear energy (2002 - 2006). The Rationale of the Pushing Robot technology comes from various considerations,more » including the need for (1) a simple and robust system, capable of moving (and potentially retrieving) on up to 40 metres (m), a 2 tonne C type package (mounted on ceramic sliding runners) inside the carbon steel sleeve constituting the liner (and rock support) of a horizontal disposal cell, (2) small annular clearances between the package and the liner, (3) compactness of the device to be transferred from surface to underground, jointly with the package, inside a shielding cask, and (4) remote controlled operations for the sake of radioprotection. The initial design, based on gripping supports, has been replaced by a 'technical variant' based on inflatable toric jacks. It was then possible, using a test bench, to check that the Pushing Robot worked properly. Steps as high as 7 mm were successfully cleared by a dummy package pushed by the Prototype.. Based on the lessons learned by ANDRA's regarding the Prototype Pushing Robot, a new Scope of Work is being written for the Contract concerning an Industrial Scale Demonstrator. The Industrial Scale Demonstration should be completed by the end of the second Quarter of 2008. (authors)« less

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2011-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

FY 2012 USED FUEL DISPOSITION CAMPAIGN TRANSPORTATION TASK REPORT ON INL EFFORTS SUPPORTING THE MODERATOR EXCLUSION CONCEPT AND STANDARDIZED TRANSPORTATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. K. Morton

2012-08-01

Following the defunding of the Yucca Mountain Project, it is reasonable to assume that commercial used fuel will remain in storage for a longer time period than initially assumed. Previous transportation task work in FY 2011, under the Department of Energy’s Office of Nuclear Energy, Used Fuel Disposition Campaign, proposed an alternative for safely transporting used fuel regardless of the structural integrity of the used fuel, baskets, poisons, or storage canisters after an extended period of storage. This alternative assures criticality safety during transportation by implementing a concept that achieves moderator exclusion (no in-leakage of moderator into the used fuelmore » cavity). By relying upon a component inside of the transportation cask that provides a watertight function, a strong argument can be made that moderator intrusion is not credible and should not be a required assumption for criticality evaluations during normal or hypothetical accident conditions of transportation. This Transportation Task report addresses the assigned FY 2012 work that supports the proposed moderator exclusion concept as well as a standardized transportation system. The two tasks assigned were to (1) promote the proposed moderator exclusion concept to both regulatory and nuclear industry audiences and (2) advance specific technical issues in order to improve American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3 rules for storage and transportation containments. The common point behind both of the assigned tasks is to provide more options that can be used to resolve current issues being debated regarding the future transportation of used fuel after extended storage.« less

Underwater characterization of control rods for waste disposal using SMOPY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.

Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performingmore » radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were performed and these demonstrate the validity of the techniques. The results are described and analysed in the context of the measurement performance of the equipment. Different casks were fully characterized using a 60 mm{sup 3} CZT detector, to determine the total activities and spatial profiles. A total activity range measurement of 1x10{sup 8} - 1x10{sup 13} Bq/cm was found to be achievable. Finally, comments are made, based on our measurements, on the ability of this equipment for performing in-situ characterisation of wastes in the harsh environments typical of fuel assembly and waste storage ponds and silos. (authors)« less

Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Triplett, C.E.

1996-12-01

This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the formmore » Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.« less

FRAUD/SABOTAGE Killing Nuclear-Reactors!!! ``Super"alloys GENERIC ENDEMIC Wigner's-Disease IN-stability!!!

NASA Astrophysics Data System (ADS)

Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

2010-03-01

Siegel [[J.Mag.Mag.Mtls.7,312(78); PSS(a)11,45(72); Semis.& Insuls.5(79)] (at: ORNL, ANS, Westin``KL"ouse, PSEG, IAEA, ABB) warning of old/new nuclear-reactors/spent-fuel-casks/refineries/ jet/missile/rocket-engines austenitic/FCC Ni/Fe-based (so MIS- called)``super"alloys(182/82;Hastelloy-X; 600;304/304L-SSs; 690 !!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's- diseas(WD)[J.Appl.Phys.17,857(46)]; Ostwald-ripening; spinodal- decomposition; overageing-embrittlement; thermomechanical- INstability: Mayo[Google: ``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: [Siegel<<<``Fert"(88) 2007-Nobel/Wolf/Japan-prizes]necessitating NRC inspections on 40+25=65 Westin``KL"ouse PWRs(12/06)]; Lai[Met.Trans.AIME,9A,827 (78)]-Sabol-Stickler[PSS(70)]; Ashpahani[Intl.Conf. H in Metals (77)]; Russell[Prog. Mtls.Sci.(83)]; Pollard[last UCS rept. (9/95)]; Lofaro[BNL/DOE/NRC Repts.]; Pringle[Nuclear-Power:From Physics to Politics(79)]; Hoffman[animatedsoftware.com],...what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrit- tlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n``u''tional-la``v''atories sabotage!!!

FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: "Super"alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

NASA Astrophysics Data System (ADS)

Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

2010-03-01

Carbides solid-state chemistry domination of old/new nuclear- reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe-based(so miscalled)``super"alloys(182/82; Hastelloy-X,600,304/304L-SSs,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-diseas(WD)[J.Appl.Phys.17,857 (1946)]/Ostwald-ripening/spinodal-decomposition/overageing- embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google:``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: Siegel[J.Mag.Mag.Mtls.7,312 (1978)]<<<``Fert"-"Gruenberg"(1988/89)2007-physics Nobel/Wolf/ Japan-prizes]necessitating NRC-inspections of 40+25 = 65 Westin- ``KLouse PWRs(12/2006)]-Lai[Met.Trans.AIME,9A,827(1978)]-Sabol- Stickler[Phys.Stat.Sol.(1970)]-Ashpahani[Intl.Conf. H in Metals, Paris(1977]-Russell[Prog.Mtls.Sci.(1983)]-Pollard[last UCS rept. (9/1995)]-Lofaro[BNL/DOE/NRC Repts.]-Pringle[Nuclear-Power:From Physics to Politics(1979)]-Hoffman[animatedsoftware.com], what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embritt- lement caused brittle-fracture cracking from early/ongoing AEC/ DOE-n"u"tional-la"v"atories sabotage!!!

FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: ``Super'' alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

NASA Astrophysics Data System (ADS)

O'Grady, Joseph; Bument, Arlden; Siegel, Edward

2011-03-01

Carbides solid-state chemistry domination of old/new nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines is austenitic/FCC Ni/Fe-based (so miscalled)"super"alloys(182/82;Hastelloy-X,600,304/304L-SSs,...690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-disease(WD) [J.Appl.Phys.17,857 (46)]/Ostwald-ripening/spinodal-decomposition/overageing-embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google: fLeaksCouldKill > ; - Siegel [ J . Mag . Mag . Mtls . 7 , 312 (78) = atflickr . comsearchonGiant - Magnotoresistance [Fert" [PRL(1988)]-"Gruenberg"[PRL(1989)] 2007-Nobel]necessitating NRC inspections on 40+25=65 Westin"KL"ouse PWRs(12/2006)]-Lai [Met.Trans.AIME, 9A,827(78)]-Sabol-Stickler[Phys.Stat.Sol.(70)]-Ashpahani[ Intl.Conf. Hydrogen in Metals, Paris(1977]-Russell [Prog.Mtls.Sci.(1983)]-Pollard [last UCS rept.(9/1995)]-Lofaro [BNL/DOE/NRC Repts.]-Pringle [ Nuclear-Power:From Physics to Politics(1979)]-Hoffman [animatedsoftware.com], what DOE/NRC MISlabels as "butt-welds" "stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrittlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n"u"tional-la"v"atories sabotage!!!

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

PubMed

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Audio Script for Information Center Transportation Display

DOE Office of Scientific and Technical Information (OSTI.GOV)

NA

2003-05-26

Can waste be transported safely to Yucca Mountain? Both the Department of Energy and the Nuclear Regulatory Commission have found that spent nuclear fuel can be shipped safely and securely. In fact, over the last 30 years there have been more than 2,700 shipments of spent nuclear fuel traveling more than 1.7 million miles, and there has never been a release of radioactive material harmful to the public or the environment--not one. Spent nuclear fuel is a solid material--it cannot leak, burn, or explode. The shipping containers, called casks, are the most robust in the transportation industry and must bemore » certified by the Nuclear Regulatory Commission. They are designed to protect public health and safety under normal and severe accident conditions. Typically, every ton of shipped spent fuel is contained within approximately 4 tons of protective shielding and structural materials. How many shipments would be made to Yucca Mountain? DOE would use mainly trains and some legal-weight trucks to move spent nuclear fuel and high-level radioactive waste to Yucca Mountain. Once the repository opens, DOE estimates and average of 130 rail shipments and 45 truck shipments per year for 24 years.« less