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

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

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...

CASKS (Computer Analysis of Storage casKS): A microcomputer based analysis system for storage cask design review. User`s manual to Version 1b (including program reference)

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

Chen, T.F.; Gerhard, M.A.; Trummer, D.J.

CASKS (Computer Analysis of Storage casKS) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent-fuel storage casks. The bulk of the complete program and this user`s manual are based upon the SCANS (Shipping Cask ANalysis System) program previously developed at LLNL. A number of enhancements and improvements were added to the original SCANS program to meet requirements unique to storage casks. CASKS is an easy-to-use system that calculates global response of storage casks to impact loads, pressure loads and thermal conditions. This provides reviewers withmore » a tool for an independent check on analyses submitted by licensees. CASKS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests.« less

78 FR 78165 - 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

... Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory... storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the...

78 FR 73456 - 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

...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to... requirements for the HI-STORM 100U part of the HI-STORM 100 Cask System and updates the thermal model and...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks... cask designs. Discussion This rule will add the Holtec HI-STORM Flood/Wind (FW) cask system to the list...

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...

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

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...

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

Casks (computer analysis of storage casks): A microcomputer based analysis system for storage cask review

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

Chen, T.F.; Mok, G.C.; Carlson, R.W.

1995-08-01

CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules: the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on themore » impact analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage casks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.« less

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...

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

CASKS (Computer Analysis of Storage Casks): A microcomputer based analysis system for storage cask review

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

Chen, T.F.; Mok, G.C.; Carlson, R.W.

1996-12-01

CASKS is a microcomputer based computer system developed by LLNL to assist the Nuclear Regulatory Commission in performing confirmatory analyses for licensing review of radioactive-material storage cask designs. The analysis programs of the CASKS computer system consist of four modules--the impact analysis module, the thermal analysis module, the thermally-induced stress analysis module, and the pressure-induced stress analysis module. CASKS uses a series of menus to coordinate input programs, cask analysis programs, output programs, data archive programs and databases, so the user is able to run the system in an interactive environment. This paper outlines the theoretical background on the impactmore » analysis module and the yielding surface formulation. The close agreement between the CASKS analytical predictions and the results obtained form the two storage asks drop tests performed by SNL and by BNFL at Winfrith serves as the validation of the CASKS impact analysis module.« 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...

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...

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

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

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

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...

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.

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...

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...

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...

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

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.

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...

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

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

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

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.

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

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

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

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...

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...

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...

Compton Dry-Cask Imaging System

ScienceCinema

None

2017-12-09

The Compton-Dry Cask Imaging Scanner is a system that verifies and documents the presence of spent nuclear fuel rods in dry-cask storage and determines their isotopic composition without moving or opening the cask. For more information about this project, visit http://www.inl.gov/rd100/2011/compton-dry-cask-imaging-system/

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

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.

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

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.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

DOE PAGES

Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

2016-06-15

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δk eff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« 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.

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

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

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...

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

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

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

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

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...

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

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

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

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

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.

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

Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

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

DeVoe, Remy R.; Robb, Kevin R.; Skutnik, Steven E.

Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basketmore » to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.« less

Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

DOE PAGES

DeVoe, Remy R.; Robb, Kevin R.; Skutnik, Steven E.

2017-07-08

Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basketmore » to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.« less

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...

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.

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.

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.

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

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 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 also 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 above- and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.« 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...

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...

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

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.

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

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.

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.

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

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.

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

Analysis of Transportation Options for Commercial Spent Fuel in the U.S.

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

Kalinina, Elena; Busch, Ingrid Karin

The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S.more » Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and associated transportation of spent nuclear fuel (SNF) high and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF)...« 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

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...

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...

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

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

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.

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

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...

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

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

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

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...

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...

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...

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

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

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.

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

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

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

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

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

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).

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.

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...

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

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

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...

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...

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...

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...

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.

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.

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.

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

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

78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0140] Draft Spent Fuel Storage and Transportation Interim... Spent Fuel Storage and Transportation Interim Staff Guidance No. 24 (SFST-ISG-24), Revision 0, ``The Use of a Demonstration Program as Confirmation of Integrity for Continued Storage of High Burnup Fuel...

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...

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 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...

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

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.

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

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

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

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

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

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

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.

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

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

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

How the University of Texas system responded to the need for interim storage of low-level radioactive waste materials.

PubMed

Emery, Robert J

2012-11-01

Faced with the prospect of being unable to permanently dispose of low-level radioactive wastes (LLRW) generated from teaching, research, and patient care activities, component institutions of the University of Texas System worked collaboratively to create a dedicated interim storage facility to be used until a permanent disposal facility became available. Located in a remote section of West Texas, the University of Texas System Interim Storage Facility (UTSISF) was licensed and put into operation in 1993, and since then has provided safe and secure interim storage for up to 350 drums of dry solid LLRW at any given time. Interim storage capability provided needed relief to component institutions, whose on-site waste facilities could have possibly become overburdened. Experiences gained from the licensing and operation of the site are described, and as a new permanent LLRW disposal facility emerges in Texas, a potential new role for the storage facility as a surge capacity storage site in times of natural disasters and emergencies is also discussed.

Plutonium storage criteria

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

Chung, D.; Ascanio, X.

1996-05-01

The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less thanmore » 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.« 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

Conceptual design statement of work for the immobilized low-activity waste interim storage facility project

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

Carlson, T.A., Fluor Daniel Hanford

1997-02-06

The Immobilized Low-Activity Waste Interim Storage subproject will provide storage capacity for immobilized low-activity waste product sold to the U.S. Department of Energy by the privatization contractor. This statement of work describes the work scope (encompassing definition of new installations and retrofit modifications to four existing grout vaults), to be performed by the Architect-Engineer, in preparation of a conceptual design for the Immobilized Low-Activity Waste Interim Storage Facility.

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...

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

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.

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

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.

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.

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

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...

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...

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

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

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.

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.

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

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

2014-12-01

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. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface ofmore » in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters. At both sites, the surface dust could be divided into fractions generated by manufacturing processes and by natural processes. The fraction from manufacturing processes consisted of variably-oxidized angular and spherical particles of stainless steel and iron, generated by machining and welding/cutting processes, respectively. Dust from natural sources consisted largely of detrital quartz and aluminosilicates (feldspars and clays) at both sites. At Hope Creek, soluble salts were dominated by sulfates and nitrates, mostly of calcium. Chloride was a trace component and the only chloride mineral observed by SEM was NaCl. Chloride surface loads measured by the Saltsmart™ sensors were very low, less than 60 mg m –2 on the canister top, and less than 10 mg m –2 on the canister sides. At Diablo Canyon, sea-salt aggregates of NaCl and Mg-SO 4, with minor K and Ca, were abundant in the dust, in some cases dominating the observed dust assemblage. Measured Saltsmart™ chloride surface loads were very low (<5 mg m –2); however, high canister surface temperatures damaged the Saltsmart™ sensors, and, in view of the SEM observations of abundant sea-salts on the package surfaces, the measured surface loads may not be valid. Moreover, the more heavily-loaded canister tops at Diablo Canyon were not sampled with the Saltsmart™ sensors. The observed low surface loads do not preclude chloride-induced stress corrosion cracking (CISCC) at either site, because (1) the measured data may not be valid for the Diablo Canyon canisters; (2) the surface coverage was not complete (for instance, the 45º offset between the outlet and inlet vents means that near-inlet areas, likely to have heavier dust and salt loads, were not sampled); and (3) CISCC has been experimentally been observed at salt loads as low as 5-8 mg/m 2. Experimental efforts at SNL to assess corrosion of interim storage canister materials include three tasks in FY14. First, a full-diameter canister mockup, made using materials and techniques identical to those used to make interim storage canisters, was designed and ordered from Ranor Inc., a cask vendor for Areva/TN. The mockup will be delivered prior to the end of FY14, and will be used for evaluating weld residual stresses and degrees of sensitization for typical interim storage canister welds. Following weld characterization, the mockup will be sectioned and provided to participating organizations for corrosion testing purposes. A test plan is being developed for these efforts. In a second task, experimental work was carried out to evaluate crevice corrosion of 304SS in the presence of limited reactants, as would be present on a dustcovered storage canister. This work tests the theory that limited salt loads will limit corrosion penetration over time, and is a continuation of work carried out in FY13. Laser confocal microscopy was utilized to assess the volume and depth of corrosion pits formed during the crevice corrosion tests. Results indicate that for the duration of the current experiments (100 days), no stifling of corrosion occurred due to limitations in the amount of reactants present at three different salt loadings. Finally, work has been carried out this year perfecting an instrument for depositing sea-salts onto metal surfaces for atmospheric corrosion testing purposes. The system uses an X-Y plotter system with a commercial airbrush, and deposition is monitored with a quartz crystal microbalance. The system is capable of depositing very even salt loadings, even at very low total deposition rates.« less

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

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.

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

NASA Astrophysics Data System (ADS)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

Absolute efficiency calibration of 6LiF-based solid state thermal neutron detectors

NASA Astrophysics Data System (ADS)

Finocchiaro, Paolo; Cosentino, Luigi; Lo Meo, Sergio; Nolte, Ralf; Radeck, Desiree

2018-03-01

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 the art of a promising low-cost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. A few configurations were studied with the GEANT4 simulation code, and the intrinsic efficiency of the corresponding detectors was calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured intrinsic 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.

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

Integrated waste management system costs in a MPC system

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

Supko, E.M.

1995-12-01

The impact on system costs of including a centralized interim storage facility as part of an integrated waste management system based on multi-purpose canister (MPC) technology was assessed in analyses by Energy Resources International, Inc. A system cost savings of $1 to $2 billion occurs if the Department of Energy begins spent fuel acceptance in 1998 at a centralized interim storage facility. That is, the savings associated with decreased utility spent fuel management costs will be greater than the cost of constructing and operating a centralized interim storage facility.

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

Used Nuclear Fuel: From Liability to Benefit

NASA Astrophysics Data System (ADS)

Orbach, Raymond L.

2011-03-01

Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

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

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...

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...

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.

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.

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...

The Time Needed to Implement the Blue Ribbon Commission Recommendation on Interim Storage - 13124

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

Voegele, Michael D.; Vieth, Donald

2013-07-01

The report of the Blue Ribbon Commission on America's Nuclear Future [1] makes a number of important recommendations to be considered if Congress elects to redirect U.S. high-level radioactive waste disposal policy. Setting aside for the purposes of this discussion any issues related to political forces leading to stopping progress on the Yucca Mountain project and driving the creation of the Commission, an important recommendation of the Commission was to institute prompt efforts to develop one or more consolidated storage facilities. The Blue Ribbon Commission noted that this recommended strategy for future storage and disposal facilities and operations should bemore » implemented regardless of what happens with Yucca Mountain. It is too easy, however, to focus on interim storage as an alternative to geologic disposal. The Blue Ribbon Commission report does not go far enough in addressing the magnitude of the contentious problems associated with reopening the issues of relative authorities of the states and federal government with which Congress wrestled in crafting the Nuclear Waste Policy Act [2]. The Blue Ribbon Commission recommendation for prompt adoption of an interim storage program does not appear to be fully informed about the actions that must be taken, the relative cost of the effort, or the realistic time line that would be involved. In essence, the recommendation leaves to others the details of the systems engineering analyses needed to understand the nature and details of all the operations required to reach an operational interim storage facility without derailing forever the true end goal of geologic disposal. The material presented identifies a number of impediments that must be overcome before the country could develop a centralized federal interim storage facility. In summary, and in the order presented, they are: 1. Change the law, HJR 87, PL 107-200, designating Yucca Mountain for the development of a repository. 2. Bring new nuclear waste legislation to the floor of the Senate, overcoming existing House support for Yucca Mountain; 3. Change the longstanding focus of Congress from disposal to storage; 4. Change the funding concepts embodied in the Nuclear Waste Policy Act to allow the Nuclear Waste fund to be used to pay for interim storage; 5. Reverse the Congressional policy not to give states or tribes veto or consent authority, and to reserve to Congress the authority to override a state or tribal disapproval; 6. Promulgate interim storage facility siting regulations to reflect the new policies after such changes to policy and law; 7. Complete already underway changes to storage and transportation regulations, possibly incorporating changes to reflect changes to waste disposal law; 8. Promulgate new repository siting regulations if the interim storage facility is to support repository development; 9. Identify volunteer sites, negotiate agreements, and get Congressional approval for negotiated benefits packages; 10. Design, License and develop the interim storage facility. The time required to accomplish these ten items depends on many factors. The estimate developed assumes that certain of the items must be completed before other items are started; given past criticisms of the current program, such an assumption appears appropriate. Estimated times for completion of individual items are based on historical precedent. (authors)« less

Cost Sensitivity Analysis for Consolidated Interim Storage of Spent Fuel: Evaluating the Effect of Economic Environment Parameters

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

Cumberland, Riley M.; Williams, Kent Alan; Jarrell, Joshua J.

This report evaluates how the economic environment (i.e., discount rate, inflation rate, escalation rate) can impact previously estimated differences in lifecycle costs between an integrated waste management system with an interim storage facility (ISF) and a similar system without an ISF.

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.

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

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

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

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

Essays in Energy Policy: The Interplay Between Risks and Incentives

NASA Astrophysics Data System (ADS)

Lordan-Perret, Rebecca Jane Bishop

My dissertation considers examples of how social, economic, and political incentives associated with energy production, distribution, and consumption increase the risk of harm to society and the environment. In the first essay, "Why America should move toward dry cask consolidated interim storage of used nuclear fuel," my co-authors and I discuss how the confluence of the U.S. Government and electricity utilities' political and economic incentives created a gridlock preventing a long-term nuclear waste disposal solution. We find that our current policies undermine the safety and security of the nuclear waste, and so, suggest a temporary, consolidated storage solution. In the second essay, "Import-Adjusted Fatality Rates for Individual OECD Countries Caused by Accidents in the Oil Energy Chain," my co-authors and I adopt a technique from the greenhouse gas accounting literature and assign CO2 emissions to the final consumer (rather than the producer) by allocating the risk - measured in fatalities - associated with oil production to the final consumer. The new assignments show that normal methods of tracking oil production impacts only capture part of the actual costs. In the third essay, "Insurgent Attacks on Energy Infrastructure and Electoral Institutions in Colombia," my co-authors and I consider the economic and political incentives that an energy resource create in a conflict environment. Our research shows that insurgents in Colombia, Las Fuerzas Armadas Revolucionarias de Colombia (FARC) and Ejercito de Liberacion Nacional (ELN), strategically time attacks on critical energy infrastructure during elections. These results are the first to quantify insurgent tactics to target critical energy infrastructure, which potentially undermine state capacity and democratic processes.

77 FR 50069 - National Oil and Hazardous Substances Pollution Contingency Plan; National Priorities List...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-20

.... Grace & Co., Inc./Wayne Interim Storage (USDOE) Superfund Site AGENCY: Environmental Protection Agency... issuing a Notice of Intent to Delete the W.R. Grace & Co., Inc./Wayne Interim Storage (USDOE) Superfund..., NY 10007-1866. Hand Delivery: U.S. EPA Superfund Records Center, Region II, 290 Broadway, 18th Floor...

Cost Implications of an Interim Storage Facility in the Waste Management System

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

Jarrell, Joshua J.; Joseph, III, Robert Anthony; Howard, Rob L

2016-09-01

This report provides an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). Specifically, the impacts of the timing of opening an ISF relative to opening a repository were analyzed to understand the potential effects on total system costs.

Lessons learned from the Siting Process of an Interim Storage Facility in Spain - 12024

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

Lamolla, Meritxell Martell

2012-07-01

On 29 December 2009, the Spanish government launched a site selection process to host a centralised interim storage facility for spent fuel and high-level radioactive waste. It was an unprecedented call for voluntarism among Spanish municipalities to site a controversial facility. Two nuclear municipalities, amongst a total of thirteen municipalities from five different regions, presented their candidatures to host the facility in their territories. For two years the government did not make a decision. Only in November 30, 2011, the new government elected on 20 November 2011 officially selected a non-nuclear municipality, Villar de Canas, for hosting this facility. Thismore » paper focuses on analysing the factors facilitating and hindering the siting of controversial facilities, in particular the interim storage facility in Spain. It demonstrates that involving all stakeholders in the decision-making process should not be underestimated. In the case of Spain, all regional governments where there were candidate municipalities willing to host the centralised interim storage facility, publicly opposed to the siting of the facility. (author)« less

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

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).

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

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 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...

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

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...

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

German Support Program for Retrieval and Safe Storage of Disused Radioactive Sealed Sources in Ukraine - 13194

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

Pretzsch, Gunter; Salewski, Peter; Sogalla, Martin

2013-07-01

The German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) on behalf of the Government of the Federal Republic of Germany supports the State Nuclear Regulatory Inspectorate of Ukraine (SNRIU) in enhancement of nuclear safety and radiation protection and strengthening of the physical protection. One of the main objectives of the agreement concluded by these parties in 2008 was the retrieval and safe interim storage of disused orphan high radioactive sealed sources in Ukraine. At present, the Ukrainian National Registry does not account all high active radiation sources but only for about 70 - 80 %. GRSmore » in charge of BMU to execute the program since 2008 concluded subcontracts with the waste management and interim storage facilities RADON at different regions in Ukraine as well with the waste management and interim storage facility IZOTOP at Kiev. Below selected examples of removal of high active Co-60 and Cs-137 sources from irradiation facilities at research institutes are described. By end of 2012 removal and safe interim storage of 12.000 disused radioactive sealed sources with a total activity of more than 5,7.10{sup 14} Bq was achieved within the frame of this program. The German support program will be continued up to the end of 2013 with the aim to remove and safely store almost all disused radioactive sealed sources in Ukraine. (authors)« less

40 CFR Appendix III to Part 265 - EPA Interim Primary Drinking Water Standards

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 26 2014-07-01 2014-07-01 false EPA Interim Primary Drinking Water Standards III Appendix III to Part 265 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Pt....

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

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

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...

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...

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...

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

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...

CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT

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

E.F. Loros

2000-06-23

The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS,more » as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building structures and space allocations. The Carrier/Cask Handling System interfaces with the Waste Handling Building Electrical System for electrical power.« less

Long-term storage facility for reactor compartments in Sayda Bay - German support for utilization of nuclear submarines in Russia

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

Wolff, Dietmar; Voelzke, Holger; Weber, Wolfgang

2007-07-01

The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted wastemore » monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on-shore concrete storage space to allow the radioactivity to decay. For transport of reactor compartments at the shipyard, at the dock and at the storage facility, hydraulic keel blocks, developed and supplied by German subcontractors, are used. In July 2006 the first stage of the reactor compartment storage facility was commissioned and the first seven reactor compartments have been delivered from Nerpa shipyard. Following transports of reactor compartments to the storage facility are expected in 2007. (authors)« less

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

Contribution of postpolymerization conditioning and storage environments to the mechanical properties of three interim restorative materials.

PubMed

Thompson, Geoffrey A; Luo, Qing

2014-09-01

Because polymer-based interim restorative materials are weak, even well-made restorations sometimes fail before the definitive restoration is ready for insertion. Therefore, knowing which fabrication procedures and service conditions affect mechanical properties is important, particularly over an extended period. The purpose of this study was to evaluate the effect of thermal treatment, surface sealing, thermocycling, storage media, storage temperature, and age on autopolymerizing poly(methylmethacrylate) and bis-acryl interim restorative materials. Outcome measures were flexural strength, Vickers surface microhardness, and impact strength. Flexural strength and microhardness of poly(methylmethacrylate) (Jet Acrylic) and 2 bis-acryl-composite resin (Protemp 3 Garant and Integrity) interim restorative materials were evaluated as affected by storage media, storage temperature, storage time, thermocycling, postpolymerization thermal treatment, or application of a surface sealer. In total, 2880 beam specimens (25×2×2 mm) were fabricated. Mechanical property analyses were made at 10 days, 30 days, 6 months, and 1 year after specimen preparation. Flexural strength was determined by using a 3-point bending test in a universal testing machine with a 1 kN load cell at a crosshead speed of 5.0 mm min(-1). Fracture specimens were recovered and used for determining Vickers microhardness. Measurements were made with a 0.1 N load and 15 second dwell time. Three microhardness measurements were made for each specimen, and the mean was used for reporting Vickers microhardness. Notched impact specimens (64×12.7×6.35 mm) were fabricated from Jet, Protemp 3 Garant, and Integrity interim restorative materials, yielding 288 impact specimens. Impact strengths were assessed at 10 days, 30 days, 6 months, and 1 year with a 2 J pendulum. The effects of the various experimental treatments were determined and rank ordered with analysis of variance, F ratios, and least square means differences Student t tests (α=.05). All experimental treatments investigated had significant effects on flexural strength, with material (P<.001) and thermocycling (P<.001) being dominant. Moreover, all experimental treatments investigated had a significant overall impact on Vickers microhardness with material (P<.001) and Palaseal glaze (P<.001) showing large effects. Material (P<.001) and age (P=.010) had a significant effect on impact strength. Mechanical properties of some interim polymeric materials can be improved by postpolymerization heat treatments or surface glazing. This procedure may extend the useful lifetime of some bis-acryl interim restorations. Copyright © 2014 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

Method of preparing nuclear wastes for tansportation and interim storage

DOEpatents

Bandyopadhyay, Gautam; Galvin, Thomas M.

1984-01-01

Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

Computerization of material test data reporting system : interim report.

DOT National Transportation Integrated Search

1973-09-01

This study was initiated to provide an integrated system of reporting, storing, and retrieving of construction and material test data using computerized (storage-retrieval) and quality control techniques. The findings reported in this interim report ...

Final Report: Characterization of Canister Mockup Weld Residual Stresses

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

Enos, David; Bryan, Charles R.

2016-12-01

Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the workmore » described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.« less

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)

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...

Storage for greater-than-Class C low-level radioactive waste

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

Beitel, G.A.

1991-12-31

EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) is actively pursuing technical storage alternatives for greater-than-Class C low-level radioactive waste (GTCC LLW) until a suitable licensed disposal facility is operating. A recently completed study projects that between 2200 and 6000 m{sup 3} of GTCC LLW will be generated by the year 2035; the base case estimate is 3250 m{sup 3}. The current plan envisions a disposal facility available as early as the year 2010. A long-term dedicated storage facility could be available in 1997. In the meantime, it is anticipated that a limited number of sealedmore » sources that are no longer useful and have GTCC concentrations of radionuclides will require storage. Arrangements are being made to provide this interim storage at an existing DOE waste management facility. All interim stored waste will subsequently be moved to the dedicated storage facility once it is operating. Negotiations are under way to establish a host site for interim storage, which may be operational, at the earliest, by the second quarter of 1993. Two major activities toward developing a long-term dedicated storage facility are ongoing. (a) An engineering study, which explores costs for alternatives to provide environmentally safe storage and satisfy all regulations, is being prepared. Details of some of the findings of that study will be presented. (b) There is also an effort under way to seek the assistance of one or more private companies in providing dedicated storage. Alternatives and options will be discussed.« 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.

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

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

CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

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

B. Gorpani

2000-06-23

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

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...

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.

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.

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

Developing a concept for a national used fuel interim storage facility in the United States

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

Lewis, Donald Wayne

2013-07-01

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a 'Monitored Retrievable Storage' facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to buildmore » a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE's goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility. (authors)« less

CARRIER PREPARATION BUILDING MATERIALS HANDLING SYSTEM DESCRIPTION DOCUMENT

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

E.F. Loros

2000-06-28

The Carrier Preparation Building Materials Handling System receives rail and truck shipping casks from the Carrier/Cask Transport System, and inspects and prepares the shipping casks for return to the Carrier/Cask Transport System. Carrier preparation operations for carriers/casks received at the surface repository include performing a radiation survey of the carrier and cask, removing/retracting the personnel barrier, measuring the cask temperature, removing/retracting the impact limiters, removing the cask tie-downs (if any), and installing the cask trunnions (if any). The shipping operations for carriers/casks leaving the surface repository include removing the cask trunnions (if any), installing the cask tie-downs (if any), installingmore » the impact limiters, performing a radiation survey of the cask, and installing the personnel barrier. There are four parallel carrier/cask preparation lines installed in the Carrier Preparation Building with two preparation bays in each line, each of which can accommodate carrier/cask shipping and receiving. The lines are operated concurrently to handle the waste shipping throughputs and to allow system maintenance operations. One remotely operated overhead bridge crane and one remotely operated manipulator is provided for each pair of carrier/cask preparation lines servicing four preparation bays. Remotely operated support equipment includes a manipulator and tooling and fixtures for removing and installing personnel barriers, impact limiters, cask trunnions, and cask tie-downs. Remote handling equipment is designed to facilitate maintenance, dose reduction, and replacement of interchangeable components where appropriate. Semi-automatic, manual, and backup control methods support normal, abnormal, and recovery operations. Laydown areas and equipment are included as required for transportation system components (e.g., personnel barriers and impact limiters), fixtures, and tooling to support abnormal and recovery operations. The Carrier Preparation Building Materials Handling System interfaces with the Cask/Carrier Transport System to move the carriers to and from the system. The Carrier Preparation Building System houses the equipment and provides the facility, utility, safety, communications, and auxiliary systems supporting operations and protecting personnel.« 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

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

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.

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."

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

75 FR 42575 - Electronic Signature and Storage of Form I-9, Employment Eligibility Verification

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-22

... Electronic Signature and Storage of Form I-9, Employment Eligibility Verification AGENCY: U.S. Immigration... published an interim final rule to permit electronic signature and storage of the Form I-9. 71 FR 34510..., or a combination of paper and electronic systems; Employers may change electronic storage systems as...

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

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

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.

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

Progress and future direction for the interim safe storage and disposal of Hanford high-level waste

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

Kinzer, J.E.; Wodrich, D.D.; Bacon, R.F.

This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the U.S. Department of Energy (DOE) and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System (TWRS) Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described. In these times ofmore » budget austerity, implementing an ongoing program that combines technical excellence and cost effectiveness is the near-term challenge. The technical initiatives and progress described in this paper are made more cost effective by DOE`s focus on work force productivity improvement, reduction of overhead costs, and reduction, integration and simplification of DOE regulations and operations requirements to more closely model those used in the private sector.« less

ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

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

B. Gorpani

2000-06-26

The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs formore » off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed into the cask unloading pool. In the cask unloading pool the DPC is removed from the cask and placed in an overpack and the DPC lid is severed and removed. Assemblies are removed from either an open cask or DPC and loaded into assembly baskets positioned in the basket staging rack in the assembly unloading pool. A method called ''blending'' is utilized to load DCs with a heat output of less than 11.8 kW. This involves combining hotter and cooler assemblies from different baskets. Blending requires storing some of the hotter fuel assemblies in fuel-blending inventory pools until cooler assemblies are available. The assembly baskets are then transferred from the basket staging rack to the assembly handling cell and loaded into the assembly drying vessels. After drying, the assemblies are removed from the assembly drying vessels and loaded into a DC positioned below the DC load port. After installation of a DC inner lid and temporary sealing device, the DC is transferred to the DC decontamination cell where the top area of the DC, the DC lifting collar, and the DC inner lid and temporary sealing device are decontaminated, and the DC is evacuated and backfilled with inert gas to prevent prolonged clad exposure to air. The DC is then transferred to the Disposal Container Handling System for lid welding. In another cask preparation and decontamination area, lids are replaced on the empty transportation casks and DPC overpacks, the casks and DPC overpacks are decontaminated, inspected, and transferred to the Carrier/Cask Handling System for shipment off-site. All system equipment is designed to facilitate manual or remote operation, decontamination, and maintenance. The system interfaces with the Carrier/Cask Handling System for incoming and outgoing transportation casks and DPCs. The system also interfaces with the Disposal Container Handling System, which prepares the DC for loading and subsequently seals the loaded DC. The system support interfaces are the Waste Handling Building System and other internal WHB support systems.« less

40 CFR 264.3 - Relationship to interim status standards.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Relationship to interim status standards. 264.3 Section 264.3 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES General § 264.3 Relationship to...

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

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

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

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

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

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

Maywood Interim Storage Site annual environmental report for calendar year 1991, Maywood, New Jersey. [Maywood Interim Storage Site

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

None

1992-09-01

This document describes the environmental monitoring program at the Maywood Interim Storage Site (MISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of MISS began in 1984 when congress added the site to the US Department of Energy's (DOE) Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a DOE program to identify and decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at MISS includesmore » sampling networks for radon and thoron concentrations in air; external gamma radiation-exposure; and total uranium, radium-226, radium-228, thorium-232, and thorium-230 concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in surface water, sediment, and groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment.« less

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

PubMed

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

RCRA, superfund and EPCRA hotline training module. Introduction to: Groundwater monitoring (40 cfr parts 264/265, subpart f) updated July 1996

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

NONE

1996-07-01

The module presents the requirements for groundwater monitoring at interim status and permitted treatment, storage, and disposal facilities (TSDFs) under the Resource Conservation and Recovery Act (RCRA). The goal of the module is to explain the standards and specific requirements for groundwater monitoring programs at interim status and permitted facilities.

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

Maywood interim storage site. Annual site environmental report, calendar year 1985

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

Not Available

1986-05-01

During 1985, the environmental monitoring program was continued at the Maywood Interim Storage Site (MISS), a US Department of Energy (DOE) facility located in the Borough of Maywood and the Township of Rochelle Park, New Jersey. The MISS is presently used for the storage of low-level radioactively contaminated soils. Monitoring results show that the MISS is in compliance with DOE concentration guides and radiation protection standards. Derived Concentration Guides (DCGs) represent the concentrations of radionuclides in air or water that would limit the radiation dose to 100 mrem/yr. The applicable guides have been revised since the 1984 environmental monitoring reportmore » was published. The guides applied in 1984 were based on a radiation protection standard of 500 mrem/yr; the guides applied for 1985 are based on a standard of 100 mrem/yr.« less

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

Stability of Beriplast P fibrin sealant: storage and reconstitution.

PubMed

Eberhard, Ulrich; Broder, Martin; Witzke, Günther

2006-04-26

This study was performed to investigate the stability of Beriplast P fibrin sealant (FS) across a range of storage conditions, both pre- and post-reconstitution. Storage stability of the FS was evaluated during long-term refrigeration (24 months) with or without interim storage at elevated temperatures (40 degrees C for 1 week and 25 degrees C for 1 and 3 months). Stability of individual FS components was assessed by measuring: fibrinogen content, Factor XIII activity (FXIII), thrombin activity and aprotinin potency. The package integrity of each component was also checked (sterility testing, moisture content and pH). Storage stability was also evaluated by testing the reconstituted product for adhesion (tearing force testing after mixing the solutions) and sterility. Reconstitution stability was evaluated following 3-months' storage, for up to 50 h post-reconstitution using the same tests as for the storage stability investigations. Pre-defined specifications were met for fibrinogen content, Factor XIII activity, and thrombin activity, demonstrating storage stability. Package integrity and the functionality and sterility of the reconstituted product were confirmed throughout. Reconstitution stability was demonstrated for up to 50 h following reconstitution, in terms of both tearing force and sterility tests. In conclusion, the storage stability of Beriplast P was demonstrated over a range of 24-month storage schedules including interim exposure to elevated temperature, and the reconstituted product was stable for up to 50 h.

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

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

Schilperoort, Daryl L.; Faulk, Darrin

2013-07-01

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried inmore » a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)« less

Taipower`s radioactive waste management program

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

Lee, B.C.C.

1996-09-01

Nuclear safety and radioactive waste management are the two major concerns of nuclear power in Taiwan. Recognizing that it is an issue imbued with political and social-economic concerns, Taipower has established an integrated nuclear backend management system and its associated financial and mechanism. For LLW, the Orchid Island storage facility will play an important role in bridging the gap between on-site storage and final disposal of LLW. Also, on-site interim storage of spent fuel for 40 years or longer will provide Taipower with ample time and flexibility to adopt the suitable alternative of direct disposal or reprocessing. In other words,more » by so exercising interim storage option, Taipower will be in a comfortable position to safely and permanently dispose of radwaste without unduly forgoing the opportunities of adopting better technologies or alternatives. Furthermore, Taipower will spare no efforts to communicate with the general public and make her nuclear backend management activities accountable to them.« 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

RCRA/UST, superfund, and EPCRA hotline training module. Introduction to: Groundwater monitoring (40 CFR parts 264/265, subpart F) updated as of July 1995

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

NONE

1995-11-01

The module presents the requirements for groundwater monitoring at interim status and permitted treatment, storage, and disposal facilities. It describes the groundwater monitoring criteria for interim status and permitted facilities. It explains monitoring well placement and outlines the three stages of the groundwater monitoring program for permitted facilities.

Leveraging Available Data to Support Extension of Transportation Packages Service Life

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

Dunn, K.; Abramczyk, G.; Bellamy, S.

Data obtained from testing shipping package materials have been leveraged to support extending the service life of select shipping packages while in nuclear materials transportation. Increasingly, nuclear material inventories are being transferred to an interim storage location where they will reside for extended periods of time. Use of a shipping package to store nuclear materials in an interim storage location has become more attractive for a variety of reasons. Shipping packages are robust and have a qualified pedigree for their performance in normal operation and accident conditions within the approved shipment period and storing nuclear material within a shipping packagemore » results in reduced operations for the storage facility. However, the shipping package materials of construction must maintain a level of integrity as specified by the safety basis of the storage facility through the duration of the storage period, which is typically well beyond the one year transportation window. Test programs have been established to obtain aging data on materials of construction that are the most sensitive/susceptible to aging in certain shipping package designs. The collective data are being used to support extending the service life of shipping packages in both transportation and storage.« less

SLUDGE TREATMENT PROJECT PHASE 1 SLUDGE STORAGE OPTIONS ASSESSMENT OF T PLANT VERSUS ALTERNATE STORAGE FACILITY

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

RUTHERFORD WW; GEUTHER WJ; STRANKMAN MR

2009-04-29

The CH2M HILL Plateau Remediation Company (CHPRC) has recommended to the U.S. Department of Energy (DOE) a two phase approach for removal and storage (Phase 1) and treatment and packaging for offsite shipment (Phase 2) of the sludge currently stored within the 105-K West Basin. This two phased strategy enables early removal of sludge from the 105-K West Basin by 2015, allowing remediation of historical unplanned releases of waste and closure of the 100-K Area. In Phase 1, the sludge currently stored in the Engineered Containers and Settler Tanks within the 105-K West Basin will be transferred into sludge transportmore » and storage containers (STSCs). The STSCs will be transported to an interim storage facility. In Phase 2, sludge will be processed (treated) to meet shipping and disposal requirements and the sludge will be packaged for final disposal at a geologic repository. The purpose of this study is to evaluate two alternatives for interim Phase 1 storage of K Basin sludge. The cost, schedule, and risks for sludge storage at a newly-constructed Alternate Storage Facility (ASF) are compared to those at T Plant, which has been used previously for sludge storage. Based on the results of the assessment, T Plant is recommended for Phase 1 interim storage of sludge. Key elements that support this recommendation are the following: (1) T Plant has a proven process for storing sludge; (2) T Plant storage can be implemented at a lower incremental cost than the ASF; and (3) T Plant storage has a more favorable schedule profile, which provides more float, than the ASF. Underpinning the recommendation of T Plant for sludge storage is the assumption that T Plant has a durable, extended mission independent of the K Basin sludge interim storage mission. If this assumption cannot be validated and the operating costs of T Plant are borne by the Sludge Treatment Project, the conclusions and recommendations of this study would change. The following decision-making strategy, which is dependent on the confidence that DOE has in the long term mission for T Plant, is proposed: (1) If the confidence level in a durable, extended T Plant mission independent of sludge storage is high, then the Sludge Treatment Project (STP) would continue to implement the path forward previously described in the Alternatives Report (HNF-39744). Risks to the sludge project can be minimized through the establishment of an Interface Control Document (ICD) defining agreed upon responsibilities for both the STP and T Plant Operations regarding the transfer and storage of sludge and ensuring that the T Plant upgrade and operational schedule is well integrated with the sludge storage activities. (2) If the confidence level in a durable, extended T Plant mission independent of sludge storage is uncertain, then the ASF conceptual design should be pursued on a parallel path with preparation of T Plant for sludge storage until those uncertainties are resolved. (3) Finally, if the confidence level in a durable, extended T Plant mission independent of sludge storage is low, then the ASF design should be selected to provide independence from the T Plant mission risk.« less

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

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

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

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

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

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.

Hazelwood Interim Storage Site annual environmental report for calendar year 1991, Hazelwood, Missouri. [Hazelwood Interim Storage Site

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

Not Available

1992-09-01

This document describes the environmental monitoring program at the Hazelwood Interim Storage Site (HISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of HISS began in 1984 when the site was assigned to the US Department of Energy (DOE) as part of the decontamination research and development project authorized by Congress under the 1984 Energy and Water Development Appropriations Act. DOE placed responsibility for HISS under the Formerly Utilized Sites Remedial Action Program (FUSRAP), a DOE program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of themore » nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at HISS includes sampling networks for radon concentrations in air; external gamma radiation exposure; and radium-226, thorium-230, and total uranium concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards and DCGs are established to protect public health and the environment.« 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

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

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.

Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: Defense waste processing facility

NASA Astrophysics Data System (ADS)

Huang, J. C.; Wright, W. V.

1982-04-01

The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built. High level waste is produced when reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The risks associated with the manufacture and interim storage of these two forms in the DWPF are compared. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information.

Container materials in environments of corroded spent nuclear fuel

NASA Astrophysics Data System (ADS)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

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

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 ...

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

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

75 FR 53353 - Notice of Availability of Final Interim Staff Guidance Document No. 25 “Pressure and Helium...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-31

... Guidance Document No. 25 ``Pressure and Helium Leakage Testing of the Confinement Boundary of Spent Fuel...: The Division of Spent Fuel Storage and Transportation (SFST) of the Office of Nuclear Materials Safety... Helium Leakage Testing of the Confinement Boundary of Spent Fuel Dry Storage Systems.'' This ISG...

An Information Storage and Retrieval System for Biological and Geological Data. Interim Report.

ERIC Educational Resources Information Center

Squires, Donald F.

A project is being conducted to test the feasibility of an information storage and retrieval system for museum specimen data, particularly for natural history museums. A pilot data processing system has been developed, with the specimen records from the national collections of birds, marine crustaceans, and rocks used as sample data. The research…

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

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

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

Functions and requirements document for interim store solidified high-level and transuranic waste

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

Smith-Fewell, M.A., Westinghouse Hanford

1996-05-17

The functions, requirements, interfaces, and architectures contained within the Functions and Requirements (F{ampersand}R) Document are based on the information currently contained within the TWRS Functions and Requirements database. The database also documents the set of technically defensible functions and requirements associated with the solidified waste interim storage mission.The F{ampersand}R Document provides a snapshot in time of the technical baseline for the project. The F{ampersand}R document is the product of functional analysis, requirements allocation and architectural structure definition. The technical baseline described in this document is traceable to the TWRS function 4.2.4.1, Interim Store Solidified Waste, and its related requirements, architecture,more » and interfaces.« less

A security vulnerabilities assessment tool for interim storage facilities of low-level radioactive wastes.

PubMed

Bible, J; Emery, R J; Williams, T; Wang, S

2006-11-01

Limited permanent low-level radioactive waste (LLRW) disposal capacity and correspondingly high disposal costs have resulted in the creation of numerous interim storage facilities for either decay-in-storage operations or longer term accumulation efforts. These facilities, which may be near the site of waste generation or in distal locations, often were not originally designed for the purpose of LLRW storage, particularly with regard to security. Facility security has become particularly important in light of the domestic terrorist acts of 2001, wherein LLRW, along with many other sources of radioactivity, became recognized commodities to those wishing to create disruption through the purposeful dissemination of radioactive materials. Since some LLRW materials may be in facilities that may exhibit varying degrees of security control sophistication, a security vulnerabilities assessment tool grounded in accepted criminal justice theory and security practice has been developed. The tool, which includes dedicated sections on general security, target hardening, criminalization benefits, and the presence of guardians, can be used by those not formally schooled in the security profession to assess the level of protection afforded to their respective facilities. The tool equips radiation safety practitioners with the ability to methodically and systematically assess the presence or relative status of various facility security aspects, many of which may not be considered by individuals from outside the security profession. For example, radiation safety professionals might not ordinarily consider facility lighting aspects, which is a staple for the security profession since it is widely known that crime disproportionately occurs more frequently at night or in poorly lit circumstances. Likewise, the means and associated time dimensions for detecting inventory discrepancies may not be commonly considered. The tool provides a simple means for radiation safety professionals to assess, and perhaps enhance in a reasonable fashion, the security of their interim storage operations. Aspects of the assessment tool can also be applied to other activities involving the protection of sources of radiation as well.

Nuclear criticality safety evaluation of the passage of decontaminated salt solution from the ITP filters into tank 50H for interim storage

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

Hobbs, D.T.; Davis, J.R.

This report assesses the nuclear criticality safety associated with the decontaminated salt solution after passing through the In-Tank Precipitation (ITP) filters, through the stripper columns and into Tank 50H for interim storage until transfer to the Saltstone facility. The criticality safety basis for the ITP process is documented. Criticality safety in the ITP filtrate has been analyzed under normal and process upset conditions. This report evaluates the potential for criticality due to the precipitation or crystallization of fissionable material from solution and an ITP process filter failure in which insoluble material carryover from salt dissolution is present. It is concludedmore » that no single inadvertent error will cause criticality and that the process will remain subcritical under normal and credible abnormal conditions.« 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.

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

Grout Isolation and Stabilization of Structures and Materials within Nuclear Facilities at the U.S. Department of Energy, Hanford Site, Summary - 12309

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

Phillips, S.J.; Phillips, M.; Etheridge, D.

2012-07-01

Per regulatory agreement and facility closure design, U.S. Department of Energy Hanford Site nuclear fuel cycle structures and materials require in situ isolation in perpetuity and/or interim physicochemical stabilization as a part of final disposal or interim waste removal, respectively. To this end, grout materials are being used to encase facilities structures or are being incorporated within structures containing hazardous and radioactive contaminants. Facilities where grout materials have been recently used for isolation and stabilization include: (1) spent fuel separations, (2) uranium trioxide calcining, (3) reactor fuel storage basin, (4) reactor fuel cooling basin transport rail tanker cars and casks,more » (5) cold vacuum drying and reactor fuel load-out, and (6) plutonium fuel metal finishing. Grout components primarily include: (1) portland cement, (2) fly ash, (3) aggregate, and (4) chemical admixtures. Mix designs for these typically include aggregate and non aggregate slurries and bulk powders. Placement equipment includes: (1) concrete piston line pump or boom pump truck for grout slurry, (2) progressive cavity and shearing vortex pump systems, and (3) extendable boom fork lift for bulk powder dry grout mix. Grout slurries placed within the interior of facilities were typically conveyed utilizing large diameter slick line and the equivalent diameter flexible high pressure concrete conveyance hose. Other facilities requirements dictated use of much smaller diameter flexible grout conveyance hose. Placement required direct operator location within facilities structures in most cases, whereas due to radiological dose concerns, placement has also been completed remotely with significant standoff distances. Grout performance during placement and subsequent to placement often required unique design. For example, grout placed in fuel basin structures to serve as interim stabilization materials required sufficient bearing i.e., unconfined compressive strength, to sustain heavy equipment yet, low breakout force to permit efficient removal by track hoe bucket or equivalent construction equipment. Further, flow of slurries through small orifice geometries of moderate head pressures was another typical design requirement. Phase separation of less than 1 percent was a typical design requirement for slurries. On the order of 30,000 cubic meters of cementitious grout have recently been placed in the above noted U.S. Department of Energy Hanford Site facilities or structures. Each has presented a unique challenge in mix design, equipment, grout injection or placement, and ultimate facility or structure performance. Unconfined compressive and shear strength, flow, density, mass attenuation coefficient, phase separation, air content, wash-out, parameters and others, unique to each facility or structure, dictate the grout mix design for each. Each mix design was tested under laboratory and scaled field conditions as a precursor to field deployment. Further, after injection or placement of each grout formulation, the material was field inspected either by standard laboratory testing protocols, direct physical evaluation, or both. (authors)« less

Platelet storage lesion in interim platelet unit concentrates: A comparison with buffy-coat and apheresis concentrates.

PubMed

Singh, Sukhi; Shams Hakimi, Caroline; Jeppsson, Anders; Hesse, Camilla

2017-12-01

Platelet storage lesion is characterized by morphological changes and impaired platelet function. The collection method and storage medium may influence the magnitude of the storage lesion. The aim of this study was to compare the newly introduced interim platelet unit (IPU) platelet concentrates (PCs) (additive solution SSP+, 40% residual plasma content) with the more established buffy-coat PCs (SSP, 20% residual plasma content) and apheresis PCs (autologous plasma) in terms of platelet storage lesions. Thirty PCs (n=10 for each type) were assessed by measuring metabolic parameters (lactate, glucose, and pH), platelet activation markers, and in vitro platelet aggregability on days 1, 4, and 7 after donation. The expression of platelet activation markers CD62p (P-selectin), CD63 (LAMP-3), and phosphatidylserine was measured using flow cytometry and in vitro aggregability was measured with multiple electrode aggregometry. Higher platelet activation and lower in vitro aggregability was observed in IPU than in buffy-coat PCs on day 1 after donation. In contrast, metabolic parameters, expression of platelet activation markers, and in vitro aggregability were better maintained in IPU than in buffy-coat PCs at the end of the storage period. Compared to apheresis PCs, IPU PCs had higher expression of activation markers and lower in vitro aggregability throughout storage. In conclusion, the results indicate that there are significant differences in platelet storage lesions between IPU, buffy-coat, and apheresis PCs. The quality of IPU PCs appears to be at least comparable to buffy-coat preparations. Further studies are required to distinguish the effect of the preparation methods from storage conditions. Copyright © 2017 Elsevier Ltd. All rights reserved.

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

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

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.« less

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

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

Cumberland, Riley M; Adeniyi, Abiodun Idowu; Howard, Rob L

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project teammore » is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.« 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.

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.

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).

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.

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

AUTOCASK (AUTOmatic Generation of 3-D CASK models). A microcomputer based system for shipping cask design review analysis

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

Gerhard, M.A.; Sommer, S.C.

1995-04-01

AUTOCASK (AUTOmatic Generation of 3-D CASK models) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for the structural analysis of shipping casks for radioactive material. Model specification is performed on the microcomputer, and the analyses are performed on an engineering workstation or mainframe computer. AUTOCASK is based on 80386/80486 compatible microcomputers. The system is composed of a series of menus, input programs, display programs, a mesh generation program, and archive programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests.

Cognitive Memory; A Computer Oriented Epistemological Approach to Information Storage and Retrieval. Interim Report, Phase I, 1 September 1967-28 February 1969.

ERIC Educational Resources Information Center

Illinois Univ., Urbana. Coordinated Science Lab.

In contrast to conventional information storage and retrieval systems in which a body of knowledge is thought of as an indexed codex of documents to which access is obtained by an appropriately indexed query, this interdisciplinary study aims at an understanding of what is "knowledge" as distinct from a "data file," how this knowledge is acquired,…

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.

Maywood Interim Storage Site: Annual environmental report for calendar year 1990, Maywood, New Jersey

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

Not Available

1991-09-01

Environmental monitoring of the US Department of Energy's (DOE) Maywood Interim Storage Site (MISS) and surrounding area began in 1984. MISS is part of the Formerly Utilized Sites Remedial Action Program (FUSRAP), a DOE program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The MISS Environmental monitoring programs was established to accommodate facility characteristics, applicable regulations, hazard potential, quantities and concentrations of materials released, extent and use of affected land and water, and localmore » public interest or concern. The environmental monitoring program at MISS includes sampling networks for radon concentrations in air; external gamma radiation exposure; and total uranium, radium-226, and thorium-232 concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in surface water, sediment, and groundwater. Monitoring results are compared with applicable Environmental Protection Agency (EPA) standards; federal, state, and local applicable or relevant and appropriate requirements (ARARs); and/or DOE derived concentration guidelines (DCGs). Environmental standards, ARARs, and DCGs are established to protect public health and the environment. Results from the 1990 environmental monitoring program show that concentrations of the contaminants of concern were all below applicable standards. Because the site is used only for interim storage and produces no processing effluents, all monitoring, except for radon and direct gamma radiation, was done on a quarterly basis. 18 refs., 17 figs., 28 tabs.« less

Preparation of actinide boride materials via solid-state metathesis reactions and actinide dicarbollide precursors

NASA Astrophysics Data System (ADS)

Lupinetti, Anthony J.; Fife, Julie; Garcia, Eduardo; Abney, Kent D.

2000-07-01

Information gaps exist in the knowledge base needed for choosing among the alternate processes to be used in the safe conversion of fissile materials to optimal forms for safe interim storage, long-term storage, and ultimate disposition. The current baseline storage technology for various wastes uses borosilicate glasses.1 The focus of this paper is the synthesis of actinide-containing ceramic materials at low and moderate temperatures (200 °C-1000 °C) using molecular and polymeric actinide borane and carborane complexes.

40 CFR 265.110 - Applicability.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Closure... through 265.115 (which concern closure) apply to the owners and operators of all hazardous waste...

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

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

Mok, G.C.; Thomas, G.R.; Gerhard, M.A.

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

Buffalo Metropolitan Area, New York Water Resources Management. Interim Report on Feasibility of Flood Management. Appendices.

DTIC Science & Technology

1983-07-01

storage areas were taken into account during the flood routings. AI.36 The computer program REVPULS, developed for this report, reverse Modified Puls...routed the hydrograph at Batavia through the storage upstream of the LVRR embankment. Subtracting this reverse -routed hydrograph from the combined...segments to form a more accurate reconstitution. The hydrographs upstream of Batavia were derived by reverse -routing and prorating by drainage area. Table

40 CFR 265.111 - Closure performance standard.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Section 265.111 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...) Controls, minimizes or eliminates, to the extent necessary to protect human health and the environment...

40 CFR 265.111 - Closure performance standard.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Section 265.111 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...) Controls, minimizes or eliminates, to the extent necessary to protect human health and the environment...

40 CFR 265.1 - Purpose, scope, and applicability.

Code of Federal Regulations, 2011 CFR

2011-07-01

....1 Section 265.1 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... establish minimum national standards that define the acceptable management of hazardous waste during the...

40 CFR 265.223 - Containment system.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 265.223 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL..., such as grass, shale, or rock, to minimize wind and water erosion and to preserve their structural...

40 CFR 265.72 - Manifest discrepancies.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 265.72 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL... waste solvent substituted for waste acid, or toxic constituents not reported on the manifest or shipping...

40 CFR 265.72 - Manifest discrepancies.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 265.72 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL... waste solvent substituted for waste acid, or toxic constituents not reported on the manifest or shipping...

40 CFR 265.72 - Manifest discrepancies.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 265.72 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL... waste solvent substituted for waste acid, or toxic constituents not reported on the manifest or shipping...

40 CFR 265.340 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES... hazardous waste incinerators (as defined in § 260.10 of this chapter), except as § 265.1 provides otherwise...

40 CFR 265.340 - Applicability.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES... hazardous waste incinerators (as defined in § 260.10 of this chapter), except as § 265.1 provides otherwise...

40 CFR 265.54 - Amendment of contingency plan.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Section 265.54 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND..., explosions, or releases of hazardous waste or hazardous waste constituents, or changes the response necessary...

40 CFR 265.228 - Closure and post-closure care.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Section 265.228 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... or operator must: (1) Remove or decontaminate all waste residues, contaminated containment system...

40 CFR 265.54 - Amendment of contingency plan.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Section 265.54 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND..., explosions, or releases of hazardous waste or hazardous waste constituents, or changes the response necessary...

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

Superfund record of decision (EPA Region 3): US Defense General Supply Center, Operable Unit 9, Chesterfield County, VA, September 1993

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

Not Available

1993-09-01

The decision document presents the selected interim remedial action for Operable Unit 9 (OU9) at the Defense General Supply Center (DGSC) in Chesterfield County, Virginia near Richmond. OU9 pertains to groundwater beneath Area 50, the Open Storage Area (OSA), and the Naitonal Guard Area (NGA). This operable unit is the third of nine operable units that are currently being addressed at the DGSC. OU9 addresses interim treatment and containment of groundwater in the upper and lower aquifers beneath Area 50, the OSA, and the NGA.

Establishing a store baseline during interim storage of waste packages and a review of potential technologies for base-lining

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

McTeer, Jennifer; Morris, Jenny; Wickham, Stephen

Interim storage is an essential component of the waste management lifecycle, providing a safe, secure environment for waste packages awaiting final disposal. In order to be able to monitor and detect change or degradation of the waste packages, storage building or equipment, it is necessary to know the original condition of these components (the 'waste storage system'). This paper presents an approach to establishing the baseline for a waste-storage system, and provides guidance on the selection and implementation of potential base-lining technologies. The approach is made up of two sections; assessment of base-lining needs and definition of base-lining approach. Duringmore » the assessment of base-lining needs a review of available monitoring data and store/package records should be undertaken (if the store is operational). Evolutionary processes (affecting safety functions), and their corresponding indicators, that can be measured to provide a baseline for the waste-storage system should then be identified in order for the most suitable indicators to be selected for base-lining. In defining the approach, identification of opportunities to collect data and constraints is undertaken before selecting the techniques for base-lining and developing a base-lining plan. Base-lining data may be used to establish that the state of the packages is consistent with the waste acceptance criteria for the storage facility and to support the interpretation of monitoring and inspection data collected during store operations. Opportunities and constraints are identified for different store and package types. Technologies that could potentially be used to measure baseline indicators are also reviewed. (authors)« less

Multi Canister Overpack (MCO) Topical Report [SEC 1 THRU 3

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

LORENZ, B.D.

In February 1995, the US Department of Energy (DOE) approved the Spent Nuclear Fuel (SNF) Project's ''Path Forward'' recommendation for resolution of the safety and environmental concerns associated with the deteriorating SNF stored in the Hanford Site's K Basins (Hansen 1995). The recommendation included an aggressive series of projects to design, construct, and operate systems and facilitates to permit the safe retrieval, packaging, transport, conditions, and interim storage of the K Basins' SNF. The facilities are the Cold VAcuum Drying Facility (CVDF) in the 100 K Area of the Hanford Site and the Canister Storage building (CSB) in the 200more » East Area. The K Basins' SNF is to be cleaned, repackaged in multi-canister overpacks (MCOs), removed from the K Basins, and transported to the CVDF for initial drying. The MCOs would then be moved to the CSB and weld sealed (Loscoe 1996) for interim storage (about 40 years). One of the major tasks associated with the initial Path Forward activities is the development and maintenance of the safety documentation. In addition to meeting the construction needs for new structures, the safety documentation for each must be generated.« less

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

40 CFR 265.1 - Purpose, scope, and applicability.

Code of Federal Regulations, 2013 CFR

2013-07-01

... waste is necessary to protect human health or the environment, that official or specialist may authorize....1 Section 265.1 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.1 - Purpose, scope, and applicability.

Code of Federal Regulations, 2012 CFR

2012-07-01

... waste is necessary to protect human health or the environment, that official or specialist may authorize....1 Section 265.1 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.1 - Purpose, scope, and applicability.

Code of Federal Regulations, 2014 CFR

2014-07-01

... waste is necessary to protect human health or the environment, that official or specialist may authorize....1 Section 265.1 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.142 - Cost estimate for closure.

Code of Federal Regulations, 2014 CFR

2014-07-01

... Section 265.142 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... salvage value that may be realized with the sale of hazardous wastes, or non-hazardous wastes if...

40 CFR 265.142 - Cost estimate for closure.

Code of Federal Regulations, 2012 CFR

2012-07-01

... Section 265.142 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... salvage value that may be realized with the sale of hazardous wastes, or non-hazardous wastes if...

40 CFR 265.142 - Cost estimate for closure.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Section 265.142 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... salvage value that may be realized with the sale of hazardous wastes, or non-hazardous wastes if...

40 CFR 265.113 - Closure; time allowed for closure.

Code of Federal Regulations, 2010 CFR

2010-07-01

... includes an amended waste analysis plan, ground-water monitoring and response program, human exposure....113 Section 265.113 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.77 - Additional reports.

Code of Federal Regulations, 2011 CFR

2011-07-01

....77 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL... submitting the biennial report and unmanifested waste reports described in §§ 265.75 and 265.76, the owner or...

40 CFR 265.76 - Unmanifested waste report.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 27 2012-07-01 2012-07-01 false Unmanifested waste report. 265.76 Section 265.76 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.1059 - Standards: Delay of repair.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Section 265.1059 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... technically infeasible without a hazardous waste management unit shutdown. In such a case, repair of this...

40 CFR 265.13 - General waste analysis.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 27 2013-07-01 2013-07-01 false General waste analysis. 265.13 Section 265.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL...

40 CFR 265.76 - Unmanifested waste report.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 26 2014-07-01 2014-07-01 false Unmanifested waste report. 265.76 Section 265.76 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.76 - Unmanifested waste report.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 26 2011-07-01 2011-07-01 false Unmanifested waste report. 265.76 Section 265.76 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.13 - General waste analysis.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 26 2011-07-01 2011-07-01 false General waste analysis. 265.13 Section 265.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL...

40 CFR 265.76 - Unmanifested waste report.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 27 2013-07-01 2013-07-01 false Unmanifested waste report. 265.76 Section 265.76 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.13 - General waste analysis.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 26 2014-07-01 2014-07-01 false General waste analysis. 265.13 Section 265.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL...

40 CFR 265.76 - Unmanifested waste report.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Unmanifested waste report. 265.76 Section 265.76 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.13 - General waste analysis.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 27 2012-07-01 2012-07-01 false General waste analysis. 265.13 Section 265.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL...

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

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, D.R.

1993-03-23

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Fail-safe storage rack for irradiated fuel rod assemblies

DOEpatents

Lewis, Donald R.

1993-01-01

A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

Reinforcement of a PMMA resin for interim fixed prostheses with silica nanoparticles.

PubMed

Topouzi, Marianthi; Kontonasaki, Eleana; Bikiaris, Dimitrios; Papadopoulou, Lambrini; Paraskevopoulos, Konstantinos M; Koidis, Petros

2017-05-01

Fractures in long span provisional/interim restorations are a common complication. Adequate fracture toughness is necessary to resist occlusal forces and crack propagation, so these restorations should be constructed with materials of improved mechanical properties. The aim of this study was to investigate the possible reinforcement of neat silica nanoparticles and trietoxyvinylsilane-modified silica nanoparticles in a PMMA resin for fixed interim restorations. Composite PMMA-Silica nanoparticles powders were mixed with PMMA liquid and compact bar shaped specimens were fabricated according to the British standard BS EN ISO 127337:2005. The single-edge notched method was used to evaluate fracture toughness (three-point bending test), while the dynamic thermomechanical properties (Storage Modulus, Loss Modulus, tanδ) of a series of nanocomposites with different amounts of nanoparticles (0.25%, 0.50%, 0.75%, 1% w.t.) were evaluated. Statistical analysis was performed and the statistically significant level was set to p<0.05. The fracture toughness of all experimental composites was remarkably higher compared to control. There was a tendency to decrease of fracture toughness, by increasing the concentration of the filler. No statistically significant differences were detected among the modified/unmodified silica nanoparticles. Dynamic mechanical properties were also affected. By increasing the silica nanoparticles content an increase in Storage Modulus was recorded, while Glass Transition Temperature was shifted at higher temperatures. Under the limitations of this in-vitro study, it can be suggested that both neat silica nanoparticles and trietoxyvinylsilane-modified silica nanoparticles, especially at low concentrations, may enhance the overall performance of fixed interim prostheses, as can effectively increase the fracture toughness, the elastic modulus and the Glass Transition Temperature of PMMA resins used in fixed provisional restorations. Copyright © 2017 Elsevier Ltd. All rights reserved.

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.

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

MERRA/AS: The MERRA Analytic Services Project Interim Report

NASA Technical Reports Server (NTRS)

Schnase, John; Duffy, Dan; Tamkin, Glenn; Nadeau, Denis; Thompson, Hoot; Grieg, Cristina; Luczak, Ed; McInerney, Mark

2013-01-01

MERRA AS is a cyberinfrastructure resource that will combine iRODS-based Climate Data Server (CDS) capabilities with Coudera MapReduce to serve MERRA analytic products, store the MERRA reanalysis data collection in an HDFS to enable parallel, high-performance, storage-side data reductions, manage storage-side driver, mapper, reducer code sets and realized objects for users, and provide a library of commonly used spatiotemporal operations that can be composed to enable higher-order analyses.

40 CFR 265.254 - Design and operating requirements.

Code of Federal Regulations, 2013 CFR

2013-07-01

....254 Section 265.254 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Waste Piles § 265.254 Design and operating requirements. The owner or operator of each...

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.

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

INTERIM EPA GUIDANCE FOR GEOSPATIAL-RELATED QUALITY ASSURANCE PROJECT PLANS

EPA Science Inventory

This guidance supplements EPA Guidance for Quality,Assurance Project Plans (EPA QA/G-5), in that the focus here is on collection and use of geospatial rather than other environmental data (e.g., strictly chemical or biological data), including unique aspects of data storage, retr...

24 CFR 55.2 - Terminology.

Code of Federal Regulations, 2014 CFR

2014-04-01

... or inoperative during flood and storm events (e.g., data storage centers, generating plants...” (§ 55.2(b)(5)). When FEMA provides interim flood hazard data, such as Advisory Base Flood Elevations... data may be used as “best available information” in accordance with Executive Order 11988. However, a...

40 CFR 265.401 - General operating requirements.

Code of Federal Regulations, 2013 CFR

2013-07-01

... equipment, the process or equipment must be equipped with a means to stop this inflow (e.g., a waste feed....401 Section 265.401 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

Spent Nuclear Fuel (SNF) Project Execution Plan

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

LEROY, P.G.

2000-11-03

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

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.

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

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.

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.

40 CFR 265.221 - Design and operating requirements.

Code of Federal Regulations, 2013 CFR

2013-07-01

... effective date of a prohibition pursuant to § 268.5 of this chapter, within this 48-month period. [50 FR... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... each replacement of an existing surface impoundment unit must install two or more liners, and a...

40 CFR 265.304 - Monitoring and inspection.

Code of Federal Regulations, 2011 CFR

2011-07-01

... for two consecutive months, the amount of liquids in the sumps must be recorded at least quarterly. If... sump until the liquid level again stays below the pump operating level for two consecutive months. (c... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.304 - Monitoring and inspection.

Code of Federal Regulations, 2012 CFR

2012-07-01

... for two consecutive months, the amount of liquids in the sumps must be recorded at least quarterly. If... sump until the liquid level again stays below the pump operating level for two consecutive months. (c... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.304 - Monitoring and inspection.

Code of Federal Regulations, 2010 CFR

2010-07-01

... for two consecutive months, the amount of liquids in the sumps must be recorded at least quarterly. If... sump until the liquid level again stays below the pump operating level for two consecutive months. (c... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.226 - Monitoring and inspection.

Code of Federal Regulations, 2012 CFR

2012-07-01

... operating level for two consecutive months, the amount of liquids in the sumps must be recorded at least... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... period. (2) After the final cover is installed, the amount of liquids removed from each leak detection...

40 CFR 265.226 - Monitoring and inspection.

Code of Federal Regulations, 2010 CFR

2010-07-01

... operating level for two consecutive months, the amount of liquids in the sumps must be recorded at least... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... period. (2) After the final cover is installed, the amount of liquids removed from each leak detection...

40 CFR 265.221 - Design and operating requirements.

Code of Federal Regulations, 2014 CFR

2014-07-01

... effective date of a prohibition pursuant to § 268.5 of this chapter, within this 48-month period. [50 FR... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... each replacement of an existing surface impoundment unit must install two or more liners, and a...

40 CFR 265.221 - Design and operating requirements.

Code of Federal Regulations, 2010 CFR

2010-07-01

... effective date of a prohibition pursuant to § 268.5 of this chapter, within this 48-month period. [50 FR... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... each replacement of an existing surface impoundment unit must install two or more liners, and a...

40 CFR 265.226 - Monitoring and inspection.

Code of Federal Regulations, 2011 CFR

2011-07-01

... operating level for two consecutive months, the amount of liquids in the sumps must be recorded at least... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... period. (2) After the final cover is installed, the amount of liquids removed from each leak detection...

40 CFR 265.304 - Monitoring and inspection.

Code of Federal Regulations, 2014 CFR

2014-07-01

... for two consecutive months, the amount of liquids in the sumps must be recorded at least quarterly. If... sump until the liquid level again stays below the pump operating level for two consecutive months. (c... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.221 - Design and operating requirements.

Code of Federal Regulations, 2012 CFR

2012-07-01

... effective date of a prohibition pursuant to § 268.5 of this chapter, within this 48-month period. [50 FR... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... each replacement of an existing surface impoundment unit must install two or more liners, and a...

40 CFR 265.112 - Closure plan; amendment of plan.

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND.... By May 19, 1981, or by six months after the effective date of the rule that first subjects a facility... description of other activities necessary during the partial and final closure periods to ensure that all...

40 CFR 265.304 - Monitoring and inspection.

Code of Federal Regulations, 2013 CFR

2013-07-01

... for two consecutive months, the amount of liquids in the sumps must be recorded at least quarterly. If... sump until the liquid level again stays below the pump operating level for two consecutive months. (c... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND...

40 CFR 265.226 - Monitoring and inspection.

Code of Federal Regulations, 2013 CFR

2013-07-01

... operating level for two consecutive months, the amount of liquids in the sumps must be recorded at least... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... period. (2) After the final cover is installed, the amount of liquids removed from each leak detection...

40 CFR 265.221 - Design and operating requirements.

Code of Federal Regulations, 2011 CFR

2011-07-01

... effective date of a prohibition pursuant to § 268.5 of this chapter, within this 48-month period. [50 FR... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... each replacement of an existing surface impoundment unit must install two or more liners, and a...

40 CFR 265.226 - Monitoring and inspection.

Code of Federal Regulations, 2014 CFR

2014-07-01

... operating level for two consecutive months, the amount of liquids in the sumps must be recorded at least... (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND... period. (2) After the final cover is installed, the amount of liquids removed from each leak detection...

34 CFR 74.34 - Equipment.

Code of Federal Regulations, 2010 CFR

2010-07-01

... following standards: (1) For equipment with a current per unit fair market value of $5000 or more, the... the cost of the original project or program to the current fair market value of the equipment. (2) If... the current fair market value of the equipment, plus any reasonable shipping or interim storage costs...

7 CFR 3019.34 - Equipment.

Code of Federal Regulations, 2010 CFR

2010-01-01

... or program to the current fair market value of the equipment. If the recipient has no need for the... the current fair market value of the equipment, plus any reasonable shipping or interim storage costs... data, including date of disposal and sales price or the method used to determine current fair market...

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].

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

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

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

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.

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.

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

The small-scale treatability study sample exemption

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

Coalgate, J.

1991-01-01

In 1981, the Environmental Protection Agency (EPA) issued an interim final rule that conditionally exempted waste samples collected solely for the purpose of monitoring or testing to determine their characteristics or composition'' from RCRA Subtitle C hazardous waste regulations. This exemption (40 CFR 261.4(d)) apples to the transportation of samples between the generator and testing laboratory, temporary storage of samples at the laboratory prior to and following testing, and storage at a laboratory for specific purposes such as an enforcement action. However, the exclusion did not include large-scale samples used in treatability studies or other testing at pilot plants ormore » other experimental facilities. As a result of comments received by the EPA subsequent to the issuance of the interim final rule, the EPA reopened the comment period on the interim final rule on September 18, 1987, and specifically requested comments on whether or not the sample exclusion should be expanded to include waste samples used in small-scale treatability studies. Almost all responders commented favorably on such a proposal. As a result, the EPA issued a final rule (53 FR 27290, July 19, 1988) conditionally exempting waste samples used in small-scale treatability studies from full regulation under Subtitle C of RCRA. The question of whether or not to extend the exclusion to larger scale as proposed by the Hazardous Waste Treatment Council was deferred until a later date. This information Brief summarizes the requirements of the small-scale treatability exemption.« less

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.

40 CFR 80.156 - Liability for violations of the interim detergent program controls and prohibitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

... base gasoline component, the detergent component, or the detergent-additized post-refinery component of... component of any post-refinery component or gasoline in the storage tank containing gasoline found to be in... evidence, that the gasoline or detergent carrier caused the violation. (2) Post-refinery component non...

40 CFR 265.442 - Design and installation of new drip pads.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 26 2011-07-01 2011-07-01 false Design and installation of new drip pads. 265.442 Section 265.442 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) INTERIM STATUS STANDARDS FOR OWNERS AND OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Drip Pads §...

40 CFR 262.212 - Making the hazardous waste determination at an on-site interim status or permitted treatment...

Code of Federal Regulations, 2011 CFR

2011-07-01

... permitted treatment, storage or disposal facility. (e) If the unwanted material is a hazardous waste, the... 40 Protection of Environment 26 2011-07-01 2011-07-01 false Making the hazardous waste... 262.212 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED...

40 CFR 262.212 - Making the hazardous waste determination at an on-site interim status or permitted treatment...

Code of Federal Regulations, 2012 CFR

2012-07-01

... permitted treatment, storage or disposal facility. (e) If the unwanted material is a hazardous waste, the... 40 Protection of Environment 27 2012-07-01 2012-07-01 false Making the hazardous waste... 262.212 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED...

40 CFR 262.212 - Making the hazardous waste determination at an on-site interim status or permitted treatment...

Code of Federal Regulations, 2014 CFR

2014-07-01

... permitted treatment, storage or disposal facility. (e) If the unwanted material is a hazardous waste, the... 40 Protection of Environment 26 2014-07-01 2014-07-01 false Making the hazardous waste... 262.212 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED...

40 CFR 262.212 - Making the hazardous waste determination at an on-site interim status or permitted treatment...

Code of Federal Regulations, 2010 CFR

2010-07-01

... permitted treatment, storage or disposal facility. (e) If the unwanted material is a hazardous waste, the... 40 Protection of Environment 25 2010-07-01 2010-07-01 false Making the hazardous waste... 262.212 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED...

40 CFR 262.212 - Making the hazardous waste determination at an on-site interim status or permitted treatment...

Code of Federal Regulations, 2013 CFR

2013-07-01

... permitted treatment, storage or disposal facility. (e) If the unwanted material is a hazardous waste, the... 40 Protection of Environment 27 2013-07-01 2013-07-01 false Making the hazardous waste... 262.212 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED...

78 FR 21980 - Aging Management of Internal Surfaces, Service Level III and Other Coatings, Atmospheric Storage...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-12

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0068] Aging Management of Internal Surfaces, Service Level... Interim Staff Guidance (LR-ISG), LR-ISG-2012-02, ``Aging Management of Internal Surfaces, Service Level... proposes to revise NRC staff-recommended aging management programs (AMP) and aging management review (AMR...

Superfund Record of Decision (EPA Region 1): Otis Air National Guard/Camp Edwards, MA. (First remedial action), May 1992. Interim report

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

Not Available

1992-05-20

The 22,000-acre Otis National Guard/Camp Edwards site is a former military vehicle maintenance facility on Cape Cod, Massachusetts, within the Massachusetts Military Reservation (MMR). The Area of Contamination Chemical Spill Area Number 4 (AOC CS-4) plume extends 11,000 feet and is located 1.1 miles from the southern boundary of MMR. Wastes and equipment handled at AOC CS-4 included oils, solvents, antifreeze, battery electrolytes, paint, and waste fuels. Additionally, the northern portion of AOC CS-4 was used as a storage yard for wastes generated by shops and laboratories operating at MMR. Liquid wastes were stored in containers or underground storage tanksmore » (USTs) in an unbermed area or deposited in USTs designated for motor gasoline. The ROD addresses OU2, the interim action for MMR AOC CS-4 ground water to prevent further down gradient migration of the contaminants. The primary contaminants of concern affecting the ground water are VOCs, including PCE and TCE.« 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

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

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

Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

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

Calmus, D.B.

1994-08-25

A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferredmore » from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length.« 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

A Computerized Classroom Language Management and Recording System for Deaf and Hard of Hearing Children.

ERIC Educational Resources Information Center

Bornstein, Harry; Casella, Vicki

This interim report describes the development of a networked computerized classroom language management and recording system to assist teachers of children who are deaf or hard-of-hearing. The system will provide storage and access capability for such information as changes in instruction, language learning progress, modifications in communication…

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...

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.

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.

Maywood Interim Storage Site annual environmental report for calendar year 1991, Maywood, New Jersey. Formerly Utilized Sites Remedial Action Program (FUSRAP)

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

Not Available

1992-09-01

This document describes the environmental monitoring program at the Maywood Interim Storage Site (MISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of MISS began in 1984 when congress added the site to the US Department of Energy`s (DOE) Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a DOE program to identify and decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation`s atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at MISS includesmore » sampling networks for radon and thoron concentrations in air; external gamma radiation-exposure; and total uranium, radium-226, radium-228, thorium-232, and thorium-230 concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in surface water, sediment, and groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment.« less

Hazelwood Interim Storage Site annual environmental report for calendar year 1991, Hazelwood, Missouri

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

Not Available

1992-09-01

This document describes the environmental monitoring program at the Hazelwood Interim Storage Site (HISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of HISS began in 1984 when the site was assigned to the US Department of Energy (DOE) as part of the decontamination research and development project authorized by Congress under the 1984 Energy and Water Development Appropriations Act. DOE placed responsibility for HISS under the Formerly Utilized Sites Remedial Action Program (FUSRAP), a DOE program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of themore » nation`s atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at HISS includes sampling networks for radon concentrations in air; external gamma radiation exposure; and radium-226, thorium-230, and total uranium concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards and DCGs are established to protect public health and the environment.« less

SWSA 6 interim corrective measures environmental monitoring: FY 1991 results

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

Clapp, R.B.; Marshall, D.S.

1992-06-01

In 1988, interim corrective measures (ICMs) were implemented at Solid Waste Storage Area (SWSA) 6 at Oak Ridge National Laboratory. The SWSA 6 site was regulated under the Resource Conservation and Recovery Act (RCRA). The ICMs consist of eight large high-density polyethylene sheets placed as temporary caps to cover trenches known to contain RCRA-regulated materials. Environmental monitoring for FY 1991 consisted of collecting water levels at 13 groundwater wells outside the capped areas and 44 wells in or near the capped areas in order to identify any significant loss of hydrologic isolation of the wastes. Past annual reports show thatmore » the caps are only partially effective in keeping the waste trenches dry and that many trenches consistently or intermittently contain water.« less

SWSA 6 interim corrective measures environmental monitoring: FY 1991 results. Environmental Restoration Program

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

Clapp, R.B.; Marshall, D.S.

1992-06-01

In 1988, interim corrective measures (ICMs) were implemented at Solid Waste Storage Area (SWSA) 6 at Oak Ridge National Laboratory. The SWSA 6 site was regulated under the Resource Conservation and Recovery Act (RCRA). The ICMs consist of eight large high-density polyethylene sheets placed as temporary caps to cover trenches known to contain RCRA-regulated materials. Environmental monitoring for FY 1991 consisted of collecting water levels at 13 groundwater wells outside the capped areas and 44 wells in or near the capped areas in order to identify any significant loss of hydrologic isolation of the wastes. Past annual reports show thatmore » the caps are only partially effective in keeping the waste trenches dry and that many trenches consistently or intermittently contain water.« less

Public Preferences Related to Radioactive Waste Management in the United States: Methodology and Response Reference Report for the 2016 Energy and Environment Survey.

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

Jenkins-Smith, Hank C.; Silva, Carol L.; Gupta, Kuhika

This report presents the questions and responses to a nationwide survey taken June 2016 to track preferences of US residents concerning the environment, energy, and radioactive waste management. A focus of the 2016 survey is public perceptions on different options for managing spent nuclear fuel, including on-site storage, interim storage, deep boreholes, general purpose geologic repositories, and geologic repositories for only defense-related waste. Highlights of the survey results include the following: (1) public attention to the 2011 accident and subsequent cleanup at the Fukushima nuclear facility continues to influence the perceived balance of risk and benefit for nuclear energy; (2)more » the incident at the Waste Isolation Pilot Plant in 2014 could influence future public support for nuclear waste management; (3) public knowledge about US nuclear waste management policies has remined higher than seen prior to the Fukushima nuclear accident and submittal of the Yucca Mountain application; (6) support for a mined disposal facility is higher than for deep borehole disposal, building one more interim storage facilities, or continued on-site storage of spent nuclear fuel; (7) support for a repository that comingles commercial and defense related waste is higher than for a repository for only defense related waste; (8) the public’s level of trust accorded to the National Academies, university scientists, and local emergency responders is the highest and the level trust accorded to advocacy organizations, public utilities, and local/national press is the lowest; and (9) the public is willing to serve on citizens panels but, in general, will only modestly engage in issues related to radioactive waste management.« less

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.

Hot conditioning equipment conceptual design report

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

Bradshaw, F.W., Westinghouse Hanford

1996-08-06

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hotmore » Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.« 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.

2015-08-31

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

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

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.

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).

Hydrazine Blending and Storage Facility, Interim Response Action Implementation. Final Safety Plan

DTIC Science & Technology

1989-08-30

operators and visitors, will be required to wear a personal hydrazine dosimeter at all times. These will be available from commercial sources and/or the Naval...suspectea or carcinogenic pocen:tal for =an. -- t, :t ha a,,-cn OSL ?:L. Is I pPm or 1.3 mg/m 3 . zo pp=n4p"NIOSI. (1978) hs a oe -nded a ceiling level

FINAL INTERIM REPORT, CANDIDATE SITES, MACHINES IN USE, DATA STORAGE AND TRANSMISSION METHODS: TESTING FEASIBILITY OF 3D ULTRASOUND DATA ACQUISITION AND RELIABILITY OF DATA RETRIEVAL FROM STORED 3D IMAGES

EPA Science Inventory

The purpose of this Work Assignment, 02-03, is to examine the feasibility of collecting transmitting, and analyzing 3-D ultrasound data in the context of a multi-center study of pregnant women. The study will also examine the reliability of measurements obtained from 3-D images< ...

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

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

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

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

Hazmat storage requires a zero-risk attitude

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

Roer, M.

It does not matter whether a company accumulates, transports, treats, stores or disposes hazardous chemicals--it is held responsible by the Environmental Protection Agency for environmental damage caused by leaks and spills. As a result, facilities must take sufficient precautions to minimize damage and avoid liability under the federal Comprehensive Environmental Response, Compensation and Liability Act, applicable state statute, Occupational Safety and Health Administration regulations, and Department of Transportation (DOT) requirements. A facility may accumulate hazardous waste onsite--without a permit or having interim status--for 90 days or less, or up to 120 days with an extension. However, certain conditions must bemore » met. Companies can determine their specific storage requirements in accordance with federal regulations and local requirements. To help these companies, various laboratories have developed procedures for examining, testing, listing and labeling hazardous materials storage lockers. A pre-examination service and accompanying approval label should provide generators and authorities with an increased level of confidence when selecting storage containment systems.« less

The Yami`s opposition to the Lanyu LLW storage installation

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

Li, K.K.; Chang, S.Y.

1993-12-31

Since 1982, the solidified low-level radioactive wastes (LLW) in Taiwan, regardless of the origins, have been sent to Lanyu for interim storage. Lanyu is a small island located 80 kilometers southeast of Taiwan. Its unique Polynesian cultural characteristics make it an attractive tourist spot. Dissatisfaction of being the commonly neglected powerless minority, in addition to the political claims from the outside environmental activists made the majority of the Lanyu residents oppose the operation of the storage facility. Approximately 80,000 drums of these wastes have been sent to Lanyu. Although the radiological monitoring results demonstrated that the current operation causes negligiblemore » impact on the environment. Accounting for the fast changing social and political situations in Taiwan today, without a good public acceptance program for both sides, the continuous operation of the Lanyu LLW storage facility until the year 2002, at which time the LLW disposal facility will be commissioned, could be in limbo.« 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

Quarterly report of RCRA groundwater monitoring data for period January 1, 1993 through March 31, 1993

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

Not Available

1993-07-01

Hanford Site interim-status groundwater monitoring projects are conducted as either background, indicator parameter evaluation, or groundwater quality assessment monitoring programs as defined in the Resource Conservation and Recovery Act of 1976 (RCRA); and Interim Status Standards for Owners and Operators of Hazardous Waste Treatment, Storage, and Disposal Facilities, as amended (40 Code of Federal Regulations [CFR] 265). Compliance with the 40 CFR 265 regulations is required by the Washington Administrative Code (WAC) 173-303. This report contains data from Hanford Site groundwater monitoring projects. This quarterly report contains data received between March 8 and May 24, 1993, which are the cutoffmore » dates for this reporting period. This report may contain not only data from the January through March quarter but also data from earlier sampling events that were not previously reported.« less

Superfund Record of Decision (EPA Region 4): USMC Camp Lejeune Military Reservation, NC. (First remedial action), September 1992. Interim report

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

Not Available

1992-09-23

The 500-acre Camp Lejeune Military Reservation is located 15 miles southeast of Jacksonville, in Onslow County, North Carolina. Within the site lies the Hadnot Point Industrial Area (HPIA), which was constructed in the late 1930's. It is composed of 75 buildings and facilities, which include gas stations, offices, storage yards, maintenance shops, and a dry cleaning plant. Several areas of the HPIA have been investigated for potential contamination attributed to Marine Corps activities and operations that resulted in a generation of potentially hazardous wastes. The ROD addresses an interim remedial action for the shallow aquifer at the HPIA to protectmore » human health from exposure to VOCs and metals. The primary contaminants of concern affecting the shallow ground water aquifer are VOCs, including benzene and TCE; and metals, including arsenic, chromium, and lead.« less

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.

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

Osmanlioglu, Ahmet Erdal

Available in abstract form only. Full text of publication follows: Naturally occurring radioactive material (NORM) in concentrated forms arises both in industry and in nature where natural radioisotopes accumulate at particular sites. Technically enhanced naturally occurring radioactive materials (TE-NORM) often occurs in an acidic environment where precipitates containing radionuclides plate out onto pipe walls, filters, tank linings, etc. Because of the radionuclides are selectively deposited at these sites, radioactivity concentration is extremely higher than the natural concentration. This paper presents characterization and related considerations of TE-NORM wastes in Turkey. Generally, accumulation conditions tend to favour the build-up of radium. Asmore » radium is highly radio-toxic, handling, treatment, storage and disposal of such material requires careful management. Turkey has the only low level waste processing and storage facility (WPSF) in Istanbul. This facility has interim storage buildings and storage area for storage of packaged radioactive waste which are containing artificial radioisotopes, but there is an increasing demand for the storage to accept bulk concentrated TE-NORM wastes from iron-steel and related industries. Most of these wastes generated from scrap metal piles which are imported from other countries. These wastes generally contain radium. (authors)« less

Calculational note for the radiological and toxicological effects of a UO{sub 3} release from the T-hopper storage pad

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

Goldberg, H.J.

1998-06-18

UO{sub 3} powder is stored at the T-hopper storage area associated with the 2714-U building in the 200 west area. The T-hopper containers and 13 drums containing this material are used to store the powder on pads immediately north of the building. An interim safety basis document (WHC,1996) was issued in 1996 for the UO{sub 3} powder storage area. In this document the isotope {sup 99}Tc was not included in the source term used to calculate the radiological consequences of a postulated release of the powder. A calculations note (HNF, 1998) was issued to remedy that deficiency. The present documentmore » is a revision to that document to reflect updated data concerning the solubility of UO{sub 3} in simulated lung fluid and to utilize more realistic powder release fractions.« less

Hybrid energy storage test procedures and high power battery project FY-1995 interim report

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

Hunt, G.L.

1995-12-01

Near the end of FY 1994, DOE provided funding and guidance to INEL for two separate but closely related tasks involving high power energy storage technology. One task was intended to develop and refine application-specific test procedures appropriate to high power energy storage devices for potential use in hybrid vehicles, including batteries, ultracapacitors, flywheels, and similar devices. The second task was intended to characterize the high power capabilities of presently available battery technologies, as well as eventually to evaluate the potential high power capabilities of advanced battery technologies such as those being developed by the USABC. Since the evaluation ofmore » such technologies is necessarily dependent to some extent on the availability of appropriate test methods, these two tasks have been closely coordinated. This report is intended to summarize the activities and results for both tasks accomplished during FY-1995.« less

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

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

Enos, David; Bryan, Charles R.

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

Interim report projects funded by EEC (European Economic Community) for offshore research

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

Parrott, M.

1978-10-01

In Nov. 1973, the EEC first adopted the principle of subsidizing work in the offshore field to improve technological development activities; since 1973, the EEC has awarded three programs (for 1974, 1975, and 1977) for 95 projects. The most recent grants, in 1977, for 40 projects at a subsidy cost of $66.8 million are tabulated, showing project category, company, estimated investment, and project description. Project categories include pipelaying, pipeline transport, underwater storage, LNG storage and transport, etc. Under the grant system, the community contributes 40Vertical Bar3< to projects on exploration and production techniques, 35Vertical Bar3< for development of production equipmentmore » and machinery, and 25-30Vertical Bar3< for technological development projects for transport and storage of hydrocarbons. According to the EEC commission, 65 projects have been submitted for a fourth program for which the total amount provided in the budget is $43.8 million.« less

Postconstruction report for the mercury tanks interim action at the Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

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

Voskuil, T.L.

1993-09-01

Three underground concrete settling tanks (tanks 2101-U, 2104-U, and 2100-U) at the Y-12 Plant on the Oak Ridge Reservation in Oak Ridge, Tennessee, contained contaminated sludges contributing mercury to the Upper East Fork Poplar Creek (UEFPC). These tanks were cleaned out as an interim action under the Comprehensive Environmental Response, Compensation, and Liability Act as part of the Reduction of Mercury in Plant Effluent subproject. Cleaning out these tanks prevented the sludge that had settled in the bottom from resuspending and carrying mercury into UEFPC. Tanks 2104-U and 2100-U were returned to service and will continue to receive effluent frommore » buildings 9201-4 and 9201-5. Tank 2101-U had been abandoned and its effluent redirected to Tank 2100-U during previous activities. This interim action permanently sealed Tank 2101-U from the storm sewer system. Upon removal of materials and completion of cleanup, inspections determined that the project`s cleanup criteria had been met. The structural integrity of the tanks was also inspected, and minor cracks identified in tanks 2101-U and 2104-U were repaired. This project is considered to have been completed successfully because it met its performance objectives as addressed in the Interim Record of Decision and the work plan: to remove the waste from the three storage tanks; to ensure that the tanks were cleaned to the levels specified; to return tanks 2100-U and 2104-U to service; to isolate Tank 2101-U permanently; and to manage the wastes in an appropriate fashion.« less

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

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.

Hazelwood Interim Storage Site: Annual site environment report, Calendar year 1985

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

Not Available

1986-11-01

The Hazelwood Interim Storage Site (HISS) is presently used for the storage of low-level radioactively contaminated soils. Monitoring results show that the HISS is in compliance with DOE Derived Concentration Guides (DCGs) and radiation protection standards. During 1985, annual average radon concentrations ranged from 10 to 23% of the DCG. The highest external dose rate at the HISS was 287 mrem/yr. The measured background dose rate for the HISS area is 99 mrem/yr. The highest average annual concentration of uranium in surface water monitored in the vicinity of the HISS was 0.7% of the DOE DCG; for /sup 226/Ra itmore » was 0.3% of the applicable DCG, and for /sup 230/Th it was 1.7%. In groundwater, the highest annual average concentration of uranium was 12% of the DCG; for /sup 226/Ra it was 3.6% of the applicable DCG, and for /sup 230/Th it was 1.8%. While there are no concentration guides for stream sediments, the highest concentration of total uranium was 19 pCi/g, the highest concentration of /sup 226/Ra was 4 pCi/g, and the highest concentration of /sup 230/Th was 300 pCi/g. Radon concentrations, external gamma dose rates, and radionuclide concentrations in groundwater at the site were lower than those measured in 1984; radionuclide concentrations in surface water were roughly equivalent to 1984 levels. For sediments, a meaningful comparison with 1984 concentrations cannot be made since samples were obtained at only two locations and were only analyzed for /sup 230/Th. The calculated radiation dose to the maximally exposed individual at the HISS, considering several exposure pathways, was 5.4 mrem, which is 5% of the radiation protection standard.« less

Formerly Utilized Sites Remedial Action Program (FUSRAP) Hazelwood Interim Storage Site annual site environmental report. Calendar year 1985. [FUSRAP

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

Not Available

1986-04-01

The Hazelwood Interim Storage Site (HISS) is presently used for the storage of low-level radioactively contaminated soils. Monitoring results show that the HISS is in compliance with DOE concentration guides and radiation protection standards. Derived Concentration Guides (DCGs) represent the concentrations of radionuclides in air or water that would limit the radiation dose to 100 mrem/y. The applicable limits have been revised since the 1984 environmental monitoring report was published. The limits applied in 1984 were based on a radiation protection standard of 500 mrem/y; the limits applied for 1985 are based on a standard of 100 mrem/y. The HISSmore » is part of the Formerly Utilized Sites Remedial Action Program (FUSRAP), a DOE program to decontaminate or otherwise control sites where low-level radioactive contamination remains from the early years of the nation's atomic energy program. To determine whether the site is in compliance with DOE standards, environmental measurements are expressed as percentages of the applicable DCG, while the calculated doses to the public are expressed as percentages of the applicable radiation protection standard. The monitoring program at the HISS measures uranium, radium, and thorium concentrations in surface water, groundwater, and sediment; radon gas concentrations in air; and external gamma radiation exposure rates. Potential radiation doses to the public are also calculated. The HISS was designated for remedial action under FUSRAP because radioactivity above applicable limits was found to exist at the site and its vicinity. Elevated levels of radiation still exist in areas where remedial action has not yet been completed.« less

Progress on plutonium stabilization

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

Hurt, D.

1996-05-01

The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE`s stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities.

Wayne Interim Storage Site environmental report for calendar year 1992, 868 Black Oak Ridge Road, Wayne, New Jersey. Formerly Utilized Sites Remedial Action Program (FUSRAP)

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

Not Available

1993-05-01

This report describes the environmental surveillance program at the Wayne Interim Storage Site (WISS) and provides the results for 1992. The fenced, site, 32 km (20 mi) northwest of Newark, New Jersey, was used between 1948 and 1971 for commercial processing of monazite sand to separate natural radioisotopes - predominantly thorium. Environmental surveillance of WISS began in 1984 in accordance with Department of Energy (DOE) Order 5400.1 when Congress added the site to DOE`s Formerly Utilized Sites Remedial Action Program (FUSRAP). The environmental surveillance program at WISS includes sampling networks for radon and thoron in air; external gamma radiation exposure;more » radium-226, radium-228, thorium-230, thorium-232, total uranium, and several chemicals in surface water and sediment; and total uranium, radium-226, radium-228, thorium-230, thorium-232, and organic and inorganic chemicals in groundwater. Monitoring results are compared with applicable Environmental Protection Agency (EPA) and state standards, DOE derived concentration guides (DCGs), dose limits, and other DOE requirements. This monitoring program assists in fulfilling the DOE policy of measuring and monitoring effluents from DOE activities and calculating hypothetical doses. Results for environmental surveillance in 1992 show that the concentrations of all radioactive and most chemical contaminants were below applicable standards.« less

Management of Low-Level Radioactive Waste from Research, Hospitals and Nuclear Medical Centers in Egypt - 13469

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

Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F.

2013-07-01

The application of radioisotopes and radiation sources in medical diagnosis and therapy is an important issue. Physicians can use radioisotopes to diagnose and treat diseases. Methods of treatment, conditioning and management of low level radioactive wastes from the use of radiation sources and radioisotopes in hospitals and nuclear medicine application, are described. Solid Radioactive waste with low-level activity after accumulation, minimization, segregation and measurement, are burned or compressed in a compactor according to the international standards. Conditioned drums are transported to the interim storage site at the Egyptian Atomic Energy Authority (EAEA) represented in Hot Labs and Waste Management Centermore » (HLWMC) for storage and monitoring. (authors)« less

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.

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

Wide Area Recovery and Resiliency Program (WARRP) Decon-13 Subject Matter Expert Meeting

DTIC Science & Technology

2012-08-14

Japan, Chernobyl , Goiania Waste Screening Workshop August 14, 2012 Edward A. Tupin Center for Radiological Emergency Response Radiation Protection...Total release -10% - 20% of releases from Chernobyl (37 PBq = 1,000,000 Curies) L~:lCl.~== ~ Wide Ar£>a Contamination ~ MEXT data as of S£>pt£>mber...and longer-tenn interim storage - disposal likely will take more time 2 1 On April 26, 1986, Unit 4 of the Chernobyl Nuclear Power Plant suffered

Analysis of Dust Samples Collected from an In-Service Interim Storage System at the Maine Yankee Nuclear Site.

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

Bryan, Charles R.; Enos, David

In July, 2016, the Electric Power Research Institute and industry partners performed a field test at the Maine Yankee Nuclear Site, located near Wiscasset, Maine. The primary goal of the field test was to evaluate the use of robots in surveying the surface of an in-service interim storage canister within an overpack; however, as part of the demonstration, dust and soluble salt samples were collected from horizontal surfaces within the interim storage system. The storage system is a vertical system made by NAC International, consisting of a steel-lined concrete overpack containing a 304 stainless steel (SS) welded storage canister. Themore » canister did not contain spent fuel but rather greater-than-class-C waste, which did not generate significant heat, limiting airflow through the storage system. The surfaces that were sampled for deposits included the top of the shield plug, the side of the canister, and a shelf at the bottom of the overpack, just below the level of the pillar on which the canister sits. The samples were sent to Sandia National Laboratories for analysis. This report summarizes the results of those analyses. Because the primary goal of the field test was to evaluate the use of robots in surveying the surface of the canister within the overpack, collection of dust samples was carried out in a qualitative fashion, using paper filters and sponges as the sampling media. The sampling focused mostly on determining the composition of soluble salts present in the dust. It was anticipated that a wet substrate would more effectively extract soluble salts from the surface that was sampled, so both the sponges and the filter paper were wetted prior to being applied to the surface of the metal. Sampling was accomplished by simply pressing the damp substrate against the metal surface for two minutes, and then removing it. It is unlikely that the sampling method quantitatively collected dust or salts from the metal surface; however, both substrates did extract a significant amount of material. The paper filters collected both particles, trapped within the cellulose fibers of the filter, and salts, while the sponges collected only the soluble salts, with very few particles. Upon delivery to Sandia, both collection media were analyzed using the same methods. The soluble salts were leached from the substrates and analyzed via ion chromatography, and insoluble minerals were analyzed by scanning electron microscopy and energy dispersive X-ray spectroscopy. The insoluble minerals were found to consist largely of terrestrially-derived mineral fragments, dominantly quartz and biotite. Large (mm-sized) aggregates of calcium carbonate, calcium silicate, and calcium aluminum silicate were also present. The aggregates had one flat, smooth surface and one well crystallized surface, and were interpreted to be efflorescence on the inside of the overpack and in the vent, formed by seepage of cement pore fluids through joints in the steel liner of the overpack. Such efflorescence was commonly observed during the boroscope inspection of the storage system at the site. The material may have flaked off and fallen to the point where the dust was collected, or may have brushed off onto the sponges when the robot was retrieved through the inlet vent. Chemical analysis showed that the soluble salts on the shield plug were Ca- and Na-rich, with lesser K and minor Mg; the anionic component was dominated by SO 4 and Cl, with minor amounts of NO 3 . The cation composition of the soluble salts from the overpack shelf and the canister surface was similar to the filter samples, but the anions differed significantly, being dominantly NO 3 with lesser Cl and only trace SO 4 . The salts appear to represent a mixture of sea-salts (probably partially converted to nitrates and sulfates by particle-gas conversion reactions) and continental salt aerosols. Ammonium, a common component in continental aerosols, was not observed and may have been lost by degassing from the canister surface or after collection during sample storage and transportation. The demonstration at Maine Yankee has shown that the robot and sampling method used for the test can successfully be used to collect soluble salts from metal surfaces within an interim storage system overpack. The results were consistent from sample to sample, suggesting that a representative sample of the soluble salts was being collected. However, it is unlikely that the salt samples collected here represent quantitative sampling of the salts on the surfaces evaluated; for that reason, chloride densities per unit area are not presented here. It should also be noted that the relevance to storage systems at the site that contain SNF may be limited, because a heat- generating canister will result in greater airflow through the overpack, affecting dust deposition rates and possibly salt compositions.« 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

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

NASA Astrophysics Data System (ADS)

Zuloaga, P.; Ordoñez, M.; Andrade, C.; Castellote, M.

2011-04-01

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Colonie Interim Storage Site environmental report for calendar year 1992, 1130 Central Avenue, Colonie, New York. Formerly Utilized Sites Remedial Action Program (FUSRAP)

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

Not Available

1993-05-01

This report describes the environmental surveillance program at the Colonie Interim Storage Site (CISS) and provides the results for 1992. The site is located in eastern New York State, approximately 6.4 km (4.0 mi) northwest of downtown Albany. From 1958 to 1984, National Lead (NL) Industries used the facility to manufacture various components from depleted and enriched uranium natural thorium. Environmental monitoring of CISS began in 1984 when Congress added, the site to the US Department of Energy`s (DOE) Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a program established to identify and decontaminate or otherwise control sites wheremore » residual radioactive materials remain from the early years of the nation`s atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental surveillance program at CISS includes sampling networks for external gamma radiation exposure and for thorium-232 and total uranium concentrations in surface water, sediment, and groundwater. Several chemical parameters are also measured in groundwater, including total metals, volatile organics, and water quality parameters. This surveillance program assists in fulfilling the DOE policy of measuring and monitoring effluents from DOE activities and calculating hypothetical doses. Results are compared with applicable Environmental Protection Agency (EPA) and New York State Department of Environmental Conservation (NYSDEC) standards, DOE derived concentration guides (DCGs), dose limits, and other DOE requirements.« less

Evaluation of hydrological cycle in the major European midlatitude river basins in the frame of the CORDEX project

NASA Astrophysics Data System (ADS)

Georgievski, Goran; Keuler, Klaus

2013-04-01

Water supply and its potential to increase social, economic and environmental risks are among the most critical challenges for the upcoming decades. Therefore, the assessment of the reliability of regional climate models (RCMs) to represent present-day hydrological balance of river basins is one of the most challenging tasks with high priority for climate modelling in order to estimate range of possible socio-economic impacts of the climate change. However, previous work in the frame of 4th IPCC AR and corresponding regional downscaling experiments (with focus on Europe and Danube river basin) showed that even the meteorological re-analyses provide unreliable data set for evaluations of climate model performance. Furthermore, large discrepancies among the RCMs are caused by internal model deficiencies (for example: systematic errors in dynamics, land-soil parameterizations, large-scale condensation and convection schemes), and in spite of higher resolution RCMs do not always improve much the results from GCMs, but even deteriorate it in some cases. All that has a consequence that capturing impact of climate change on hydrological cycle is not an easy task. Here we present state of the art of RCMs in the frame of the CORDEX project for Europe. First analysis shows again that even the up to date ERA-INTERIM re-analysis is not reliable for evaluation of hydrological cycle in major European midlatitude river basins (Seine, Rhine, Elbe, Oder, Vistula, Danube, Po, Rhone, Garonne and Ebro). Therefore, terrestrial water storage, a quasi observed parameter which is a combination of river discharge (from Global River Discharge Centre data set) and atmospheric moisture fluxes from ERA-INTERIM re-analysis, is used for verification. It shows qualitatively good agreement with COSMO-CLM (CCLM) regional climate simulation (abbreviated CCLM_eval) at 0.11 degrees horizontal resolution forced by ERA-INTERIM re-analysis. Furthermore, intercomparison of terrestrial water storage seasonal cycle averaged in Danube river basin for the ten years (1990-1999) overlapping period between CCLM historical experiment (abbreviated CCLM_hist), its forcing GCM (MPI-ESM-LR, here abbreviated MPI_hist) and CCLM_eval is performed. It reveals that CCLM_hist simulation is in better agreement with quasi observed terrestrial water storage than MPI_hist and CCLM_eval. This result seems promising for the assessment of impact of climate change on hydrological cycle. However, evaluation of the whole ensemble of regional climate downscaling experiments participated in CORDEX-Europe project would provide a more robust estimate.

Evaluation of Hose in Hose Transfer Line Service Life for Hanfords Interim Stabilization Program

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

TORRES, T.D.

RPP-6153, Engineering Task Plan for Hose-in-Hose Transfer System for the Interim Stabilization Program (Torres, 2000a), defines the programmatic goals, functional requirements, and technical criteria for the development and subsequent installation of waste transfer line equipment to support Hanford's Interim Stabilization Program. RPP-6028, Specification for Hose in Hose Transfer Lines for Hanford's Interim Stabilization Program (Torres, 2000b), has been issued to define the specific requirements for the design, manufacture, and verification of transfer line assemblies for specific waste transfer applications associated with Interim Stabilization. Included in RPP-6028 are tables defining the chemical constituents of concern to which transfer lines will bemore » exposed. Current Interim Stabilization Program planning forecasts that the at-grade transfer lines will be required to convey pumpable waste for as much as three years after commissioning, RPP-6028 Section 3.2.7. Performance Incentive Number ORP-05 requires that all the Single Shell Tanks be Interim Stabilized by September 30, 2003. The Tri-Party Agreement (TPA) milestone M-41-00, enforced by a federal consent decree, requires all the Single Shell Tanks to be Interim stabilized by September 30, 2004. By meeting the Performance Incentive the TPA milestone is met. Prudent engineering dictates that the equipment used to transfer waste have a life in excess of the forecasted operational time period, with some margin to allow for future adjustments to the planned schedule. This document evaluates the effective service life of the Hose-in-Hose Transfer Lines, based on information submitted by the manufacturer, published literature and calculations. The effective service life of transfer line assemblies is a function of several factors. Foremost among these are the hose material's resistance to the harmful effects of process fluid characteristics, ambient environmental conditions, exposure to ionizing radiation and the manufacturer's stated shelf life. In order to determine the transfer line service life this evaluation examines the certification of shelf life, the certification of chemical compatibility with waste, catalog information of ambient ratings and published literature on the effects of exposure to ionizing radiation on the mechanical properties of elastomeric materials. During initial hose procurements, the hose-in-hose transfer line vendor River Bend Hose Specialty (RBHS) submitted a letter, dated 6/8/00, which recommended the service and shelf life of the hose to be seven years. In submittals for later hose procurements, RBHS submitted a letter, dated 11/6/00, which recommended the service life of the hose to be three years. This submittal was followed by documentation, on 2/14/01, which submitted new storage requirements and restated the seven year shelf life. RBHS revised their original hose service life estimate to a more conservative three years due to concerns over the effects of chemicals in transferred waste. The above mentioned submittals from RBHS are the primary drivers of the three year service life limit established by this document.« 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

Uranium Enrichment Safeguards

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

Demuth, Scott F.; Trahan, Alexis Chanel

2017-06-26

DIV of facility layout, material flows, and other information provided in the DIQ. Material accountancy through an annual PIV and a number of interim inventory verifications, including UF6 cylinder identification and counting, NDA of cylinders, and DA on a sample collection of UF6. Application of C/S technologies utilizing seals and tamper-indicating devices (TIDs) on cylinders, containers, storage rooms, and IAEA instrumentation to provide continuity of knowledge between inspection. Verification of the absence of undeclared material and operations, especially HEU production, through SNRIs, LFUA of cascade halls, and environmental swipe sampling