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Sample records for fuel-storage basin superstructure

  1. Corrosion in ICPP fuel storage basins

    International Nuclear Information System (INIS)

    Dirk, W.J.

    1993-09-01

    The Idaho Chemical Processing Plant currently stores irradiated nuclear fuel in fuel storage basins. Historically, fuel has been stored for over 30 years. During the 1970's, an algae problem occurred which required higher levels of chemical treatment of the basin water to maintain visibility for fuel storage operations. This treatment led to higher levels of chlorides than seen previously which cause increased corrosion of aluminum and carbon steel, but has had little effect on the stainless steel in the basin. Corrosion measurements of select aluminum fuel storage cans, aluminum fuel storage buckets, and operational support equipment have been completed. Aluminum has exhibited good general corrosion rates, but has shown accelerated preferential attack in the form of pitting. Hot dipped zinc coated carbon steel, which has been in the basin for approximately 40 years, has shown a general corrosion rate of 4 mpy, and there is evidence of large shallow pits on the surface. A welded Type 304 stainless steel corrosion coupon has shown no attack after 13 years exposure. Galvanic couples between carbon steel welded to Type 304 stainless steel occur in fuel storage yokes exposed to the basin water. These welded couples have shown galvanic attack as well as hot weld cracking and intergranular cracking. The intergranular stress corrosion cracking is attributed to crevices formed during fabrication which allowed chlorides to concentrate

  2. 100KE/KW fuel storage basin surface volumetric factors

    International Nuclear Information System (INIS)

    Conn, K.R.

    1996-01-01

    This Supporting Document presents calculations of surface Volumetric factors for the 100KE and 100KW Fuel Storage Basins. These factors relate water level changes to basin loss or additions of water, or the equivalent water displacement volumes of objects added to or removed from the basin

  3. Control of corrosion in an aqueous nuclear fuel storage basin

    International Nuclear Information System (INIS)

    Zimmerman, C.A.

    1981-01-01

    Observations made during thirty years of experience in operating a nuclear fuel storage basin, used for storing a wide variety of spent nuclear fuels underwater have identified several forms of corrosion such as galvanic, pitting and crevice attack. Examples of some of the forms of corrosion observed and their causes are discussed, along with the measures taken to mitigate the corrosive attack. The paper also describes the procedure used to reduce corrosion by: surveillance of design, selection of materials for application in the basin, and inspection of items in the storage basin

  4. 105-N Fuel Storage Basin dewatering conceptual plan

    International Nuclear Information System (INIS)

    Schilperoort, D.L.

    1996-11-01

    This conceptual plan discusses the processes that will be used for draining and disposing of water from the 105-N Fuel Storage Basin (N Basin), and includes a description of the activities to control surface contamination and potential high dose rates encountered during dewatering. The 105-N Fuel Storage Basin is located in the 100-N Area of the Hanford Site in Richland, Washington. The processes for water disposal include water filtration, water sampling and analysis, tanker loading and unloading, surface decontamination and sealing, and clean out and disposal of residual debris and sediments during final pumping to remove the N Basin water. Water disposal is critical for the deactivation of N Reactor. A Memorandum of Understanding (MOU) between the US Department of Energy (DOE) Environmental Restoration (ER) Program and DOE Waste Management (WM) Program establishes the 200 East Effluent Treatment Facility (ETF) as the final treatment and disposal site for N Basin water and identifies pre-treatment requirements. This MOU states that water delivery will be completed no later than October 31, 1996, and will require a revision due to the current de-watering schedule date. The current MOU requires four micron filtration prior to shipment to ETF. The MOU revision for delivery date extension seeks to have the filtration limit increased to five microns, which eliminates the need for a second filter system and simplifies dewatering. For the purposes of this plan, it will be assumed that five micron filtration will be used

  5. Leak testing fuel stored in the ICPP fuel storage basin

    International Nuclear Information System (INIS)

    Lee, J.L.; Rhodes, D.W.

    1977-06-01

    Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

  6. Spent LWR fuel storage costs: reracking, AR basins, and AFR basins

    International Nuclear Information System (INIS)

    1980-01-01

    Whenever possible, fuel storage requirements will be met by reracking existing reactor basins and/or transfer of fuel to available space in other reactor basins. These alternatives represent not only the lowest cost storage options but also the most timely. They are recognized to face environmental and regulatory obstacles. However, such obstacles should be less severe than those that would be encountered with AR or AFR basin storage. When storage requirements cannot be met by the first two options, the least costly alternative for most utilities will be use of a Federal AFR. Storage costs of $100,000 to $150,000 MTU at a AFR are less costly than charges of up to $320,000/MTU that could be incurred by the use of AR basins. AFR storage costs do not include transportation from the reactor to the AFR. This cost would be paid by the utility separately. Only when a utility requires annual storage capacity for 100 MTU of spent fuel can self-storage begin to compete with AFR costs. The large reactor complexes discharging these fuel quantities are not currently those that require relief from fuel storage problems

  7. ALARA Review for the Sediment Relocation and Removal from the 105-N Fuel Storage Basin

    International Nuclear Information System (INIS)

    Demers, J.W.

    1998-01-01

    This as low as reasonable achievable (ALARA) review revision provides a description of the engineering and administrative controls used to manage personnel exposure, control contamination levels, and airborne radioactivity concentrations during sediment relocation and removal in the 105-N Fuel Storage Basin. This document updates and supercedes the ALARA review of the sediment-related activities contained in 100-N Basin Stabilization Project As Low As Reasonably Achievable Plan (BHI 1995)

  8. Data compilation report: Gas and liquid samples from K West Basin fuel storage canisters

    International Nuclear Information System (INIS)

    Trimble, D.J.

    1995-01-01

    Forty-one gas and liquid samples were taken from spent fuel storage canisters in the K West Basin during a March 1995 sampling campaign. (Spent fuel from the N Reactor is stored in sealed canisters at the bottom of the K West Basin.) A description of the sampling process, gamma energy analysis data, and quantitative gas mass spectroscopy data are documented. This documentation does not include data analysis

  9. Refinishing contamination floors in Spent Nuclear Fuels storage basins

    International Nuclear Information System (INIS)

    Huang, F.F.; Moore, F.W.

    1997-01-01

    The floors of the K Basins at the Hanford Site are refinished to make decontamination easier if spills occur as the spent nuclear fuel (SNF) is being unloaded from the basins for shipment to dry storage. Without removing the contaminated existing coating, the basin floors are to be coated with an epoxy coating material selected on the basis of the results of field tests of several paint products. The floor refinishing activities must be reviewed by a management review board to ensure that work can be performed in a controlled manner. Major documents prepared for management board review include a report on maintaining radiation exposure as low as reasonably achievable, a waste management plan, and reports on hazard classification and unreviewed safety questions. To protect personnel working in the radiation zone, Operational Health Physics prescribed the required minimum protective methods and devices in the radiological work permit. Also, industrial hygiene safety must be analyzed to establish respirator requirements for persons working in the basins. The procedure and requirements for the refinishing work are detailed in a work package approved by all safety engineers. After the refinishing work is completed, waste materials generated from the refinishing work must be disposed of according to the waste management plan

  10. Decommissioning of the 105-F and 105-H fuel storage basin in the 100 Area at the Hanford Site

    International Nuclear Information System (INIS)

    Griffin, P.W.

    1991-09-01

    The US Department of Energy (DOE) owns the eight surplus production reactors at the Hanford Site north of Richland, Washington. The fuel storage basins at the 105-F and 105-H reactors were filled with equipment, associated with the operation of the basins and clean fill in 1970. This was done to stabilize the residual sediment and a few feet of water in the reactors' fuel storage basins. This project investigates the subject basins to locate and remove overlooked fuel elements left in the basins

  11. Spent fuel consolidation in the 105KW Building fuel storage basin

    International Nuclear Information System (INIS)

    Johnson, B.H.

    1994-01-01

    This study is one element of a larger engineering study effort by WHC to examine the feasibility of irradiated fuel and sludge consolidation in the KW Basin in response to TPA Milestone (target date) M-34-00-T03. The study concludes that up to 11,500 fuel storage canisters could be accommodated in the KW Basin with modifications. These modifications would include provisions for multi-tiered canister storage involving the fabrication and installation of new storage racks and installation of additional decay heat removal systems for control of basin water temperature. The ability of existing systems to control radionuclide concentrations in the basin water is examined. The study discusses requirements for spent nuclear fuel inventory given the proposed multi-tiered storage arrangement, the impact of the consolidated mass on the KW Basin structure, and criticality issues associated with multi-tiered storage

  12. Corrosion of assemblies in fuel-storage basins at Savannah River Plant

    International Nuclear Information System (INIS)

    Wollam, C.D.

    1980-09-01

    Pitting of reactor assemblies has been the major corrosion problem in the Savannah River Plant fuel storage basins. From 1972 to 1976 many reactor assemblies experienced severe pitting corrosion with rates up to 9.3 mm/y. Poor cladding, high concentrations of iron and chloride ions in the water, a galvanic couple between the aluminum clad assemblies and the stainless steel hangers, and scratches in the oxide layer on assemblies have been identified as contributors to the problem. This paper describes the examinations and tests that were conducted and discusses a theory that explains the observed phenomena

  13. Review of ALARA plan for activities at the 105 K-East fuel storage basin

    International Nuclear Information System (INIS)

    Vargo, G.J.; Durham, J.S.; Hickey, E.E.; Stansbury, P.S.; Cicotte, G.R.

    1994-09-01

    As part of its ongoing efforts to reduce doses to workers to levels as low as reasonably achievable (ALARA), Westinghouse Hanford Company (WHC) tasked the Health Protection Department of the Pacific Northwest Laboratory (PNL) to review operations at the 105 K-East Fuel Storage Basin (105 K-East). This review included both routine operations and a proposed campaign to encapsulate N-Reactor fuel stored there. This report summarizes the results of PNL's reviews of policy, procedures, and practices for operations at 105 K-East as well as an evaluation of the major sources of occupational radiation exposures. Where possible, data previously collected by WHC and its predecessors were used. In addition, PNL staff developed a three-dimensional model of the radiological environment within 105 K-East to assess the relative contributions of different radiation sources to worker dose and to provide a decision tool for use in evaluating alternative methods of dose rate reduction. The model developed by PNL indicates that for most areas in the basin the primary source of occupational radiation exposure is the contaminated concrete surfaces of the basin near the waterline. Basin cooling water piping represents a significant source in a number of areas, particularly the Technical Viewing Pit. This report contains specific recommendations to reduce the impact of these sources of occupational radiation exposure in 105 K-East. Other recommendations to reduce doses to workers during activities such as filter changes and filter sampling are also included

  14. CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

    International Nuclear Information System (INIS)

    JENSEN, M.A.

    2005-01-01

    In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records will exist and that it will be increasingly difficult to immediately characterize the nature of the suspect fissionable

  15. Inspection and Analysis of Aluminum Storage Racks in Spent Fuel Storage Basins

    International Nuclear Information System (INIS)

    Howell, J.P.; Nelson, D.Z.

    1998-07-01

    Aluminum-clad spent nuclear fuel is stored in water filled basins at the Savannah River Site awaiting processing or other disposition. After more than 35 years of service underwater, the aluminum storage racks that position the fuel bundles in the basin were replaced. During the removal of the racks from the basin, a failure occurred in one of the racks and the Savannah River Technology Center was asked to investigate. This paper presents results of the failure analysis and provides a discussion of the effects of corrosion on the structural integrity of the storage racks

  16. Alternative concepts for spent fuel storage basin expansion at Morris Operation

    International Nuclear Information System (INIS)

    Graf, W.A. Jr.; King, C.E.; Miller, G.P.; Shadel, F.H.; Sloat, R.J.

    1980-08-01

    Alternative concepts for increasing basin capabilities for storage of spent fuel at the Morris Operation have been defined in a series of simplified flow diagrams and equipment schematics. Preliminary concepts have been outlined for (1) construction alternatives for an add-on basin, (2) high-density baskets for storage of fuel bundles or possible consolidated fuel rods in the existing or add-on basins, (3) modifications to the existing facility for increasing cask handling and fuel receiving capabilities and (4) accumulation, treatment and disposal of radwastes from storage operations. Preliminary capital and operating costs have been prepared and resource and schedule requirements for implementing the concepts have been estimated. The basin expansion alternatives would readily complement potential dry storage projects at the site in an integrated multi-stage program that could provide a total storage capacity of up to 7000 tonnes of spent fuel

  17. Spent Fuel and Waste Management Activities for Cleanout of the 105 F Fuel Storage Basin at Hanford

    International Nuclear Information System (INIS)

    Morton, M. R.; Rodovsky, T. J.; Day, R. S.

    2002-01-01

    Clean-out of the F Reactor fuel storage basin (FSB) by the Environmental Restoration Contractor (ERC) is an element of the FSB decontamination and decommissioning and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the F Reactor FSB was reduced to approximately 0.6 m (2 ft) over the floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6 m (20 ft) of local surface material (essentially a fine sand). The reactor FSB backfill cleanout involves the potential removal of spent nuclear fuel (SNF) that may have been left in the basin unintentionally. Based on previous cleanout of four water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980s), it was estimated that up to five SNF elements could be discovered in the F Reactor FSB (1). In reality, a total of 10 SNF elements have been found in the first 25% of the F Reactor FSB excavation. This paper discusses the technical and programmatic challenges of performing this decommissioning effort with some of the controls needed for SNF management. The paper also highlights how many various technologies were married into a complete package to address the issue at hand and show how no one tool could be used to complete the job; but by combining the use of multiple tools, progress is being made

  18. Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1979-09-01

    A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems

  19. Determination of uranium metal concentration in irradiated fuel storage basin sludge using selective dissolution

    International Nuclear Information System (INIS)

    Delegard, C.H.; Sinkov, S.I.; Chenault, J.W.; Schmidt, A.J.; Pool, K.N.; Welsh, T.L.

    2014-01-01

    Irradiated uranium metal fuel was stored underwater in the K East and K West storage basins at the US Department of Energy Hanford Site. The uranium metal under damaged cladding reacted with water to generate hydrogen gas, uranium oxides, and spalled uranium metal particles which intermingled with other particulates to form sludge. While the fuel has been removed, uranium metal in the sludge remains hazardous. An expeditious routine method to analyze 0.03 wt% uranium metal in the presence of >30 wt% total uranium was needed to support safe sludge management and processing. A selective dissolution method was designed based on the rapid uranium oxide dissolution but very low uranium metal corrosion rates in hot concentrated phosphoric acid. The uranium metal-bearing heel from the phosphoric acid step then is rinsed before the uranium metal is dissolved in hot concentrated nitric acid for analysis. Technical underpinnings of the selective dissolution method, including the influence of sludge components, were investigated to design the steps and define the reagents, quantities, concentrations, temperatures, and times within the selective dissolution analysis. Tests with simulant sludge proved the technique feasible. Tests with genuine sludge showed a 0.0028 ± 0.0037 wt% (at one standard deviation) uranium metal analytical background, a 0.011 wt% detection limit, and a 0.030 wt% quantitation limit in settled (wet) sludge. In tests using genuine K Basin sludge spiked with uranium metal at concentrations above the 0.030 wt% ± 25 % (relative) quantitation limit, uranium metal recoveries averaged 99.5 % with a relative standard deviation of 3.5 %. (author)

  20. Assessment of the requirements for placing and maintaining Savannah River Site spent fuel storage basins under International Atomic Energy Agency safeguards

    International Nuclear Information System (INIS)

    Amacker, O.P. Jr.; Curtis, M.M.; Delegard, C.H.; Hsue, S.T.; Whitesel, R.N.

    1997-03-01

    The United States is considering the offer of irradiated research reactor spent fuel (RRSF) for international safeguards applied by the International Atomic Energy Agency (IAEA). The offer would be to add one or more spent fuel storage basins to the list of facilities eligible for IAEA safeguards. The fuel to be safeguarded would be stored in basins on the Savannah River Site (SRS). This RRSF potentially can include returns of Material Test Reactor (MTR) VAX fuel from Argentina, Brazil, and Chile (ABC); returns from other foreign research reactors; and fuel from domestic research reactors. Basins on the SRS being considered for this fuel storage are the Receiving Basin for Offsite Fuel (RBOF) and the L-Area Disassembly Basin (L-Basin). A working group of SRS, U.S. Department of Energy International Safeguards Division (NN-44), and National Laboratory personnel with experience in IAEA safeguards was convened to consider the requirements for applying the safeguards to this material. The working group projected the safeguards requirements and described alternatives

  1. Fire barrier evaluation of the wall between spent nuclear fuel storage basins and reactor areas, 105KE and 105KW

    International Nuclear Information System (INIS)

    Johnson, B.H.

    1994-01-01

    This evaluation is intended to provide a comprehensive assessment of the risks from fire and fire related perils for the fire barriers between spent nuclear fuel storage basins and reactor areas, 105KE and 105KW. As a result of this fire barrier evaluation the present walls and the components thereof are not a true listed fire rated assembly. However, due to the construction of these barriers and the components thereof, these barriers will provide an equivalent level of protection provided the recommendations in Section 8.0 of this report are completed. These recommended upgrades are based upon sound engineering practice by a Registered Fire Protection Engineer. The construction of the barrier are substantial enough to provide the required 2-hr fire resistance rating. The primary concern is the numerous penetrations in the barrier. There are many penetrations that are adequate and no additional work is required. These penetrations are the ones that were poured-in-place at the time of construction. The penetrations that are of concern are some of the doors, the HVAC ducts, and the unsealed piping and conduit penetrations. There are several metal doors that should be replaced because the existing doors have either a non-approved window or louver that will not limit the spread of fire to one side of the barrier. All unsealed piping and conduit penetrations should be firestopped with an approved firestopping material. The existing non-active ducts that pass through the barrier should be disconnected at the barrier and the opening sealed with an approved firestopping method

  2. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    International Nuclear Information System (INIS)

    1995-07-01

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

  3. DEMOLISHING A COLD-WAR-ERA FULE-STORAGE BASIN SUPERSTRUCTURE LADEN WITH ASBESTOS

    International Nuclear Information System (INIS)

    LLOYD, E.R.; STEVENS, J.M.; DAGAN, E.B.; ORGILL, T.K.; GREEN, M.A.; LARSON, C.H.; ZINSLI, L.C.

    2009-01-01

    The K East (KE) Basin facilities are located near the north end of the Hanford Site's 100 K area. The facilities were built in 1950 as part of the KE Reactor complex and constructed within 400 meters of the Columbia River, which is the largest river in the Pacific Northwest and by volume the fourth largest river in the United States. The basin, located adjacent to the reactor, was used for the underwater storage of irradiated nuclear fuel discharged from the reactor. The basin was covered by a superstructure comprising steel columns and beams, concrete, and cement asbestos board (CAB) siding. The project's mission was to complete demolition of the structure over the KE Basin within six months of turnover from facility deactivation activities. The demolition project team applied open-air demolition techniques to bring the facility to slab-on-grade. Several innovative techniques were used to control contamination and maintain contamination control within the confines of the demolition exclusion zone. The techniques, which focused on a defense-in-depth approach, included spraying fixatives on interior and exterior surfaces before demolition began; applying fixatives during the demolition; misting using a fine spray of water during demolition; and demolishing the facility systematically. Another innovative approach that made demolition easier was to demolish the building with the non-friable CAB remaining in place. The CAB siding covered the exterior of the building and portions of the interior walls, and was an integral part of the multiple-layered roof. The project evaluated the risks involved in removing the CAB material in a radiologically contaminated environment and determined that radiological dose rates and exposure to radiological contamination and industrial hazards would be significantly reduced by using heavy equipment to remove the CAB during demolition. The ability to perform this demolition safely and without spreading contamination (radiological or

  4. LESSONS LEARNED FROM CLEANING OUT THE SLUDGE FROM THE SPENT FUEL STORAGE BASINS AT HANFORD ICEM-07

    Energy Technology Data Exchange (ETDEWEB)

    KNOLLMEYER PM

    2007-08-31

    Until 2004, the K Basins at Hanford, in southeastern Washington State, held the largest collection of spent nuclear fuel in the United States Department of Energy (DOE) complex. The K East and K West Basins are massive pools each holding more than 4 million liters of water - that sit less than 450 meters from the Columbia River. In a significant multi-year campaign that ended in 2004, Fluor Hanford removed all of the fuel from the two Basins, over 2,300 metric tons (4.6 million pounds), dried it, and then placed it into dry storage in a specially designed facility away from the River. Removing the fuel, however, did not finish the cleanup work at the K Basins. The years of underwater storage had corroded the metallic uranium fuel, leaving behind a thick and sometimes hard-packed layer of sludge that coated the walls, floors and equipment inside the Basins. In places, the depth of the sludge was measured in feet rather than inches, and its composition was definitely not uniform. Together the Basins held an estimated 50 cubic meters of sludge (42 cubic meters in K East and 8 cubic meters in K West). The K East sludge retrieval and transfer work was completed in May 2007. Vacuuming up the sludge into large underwater containers in each of the Basins and then consolidating it all in containers in the K West Basin have presented significant challenges, some unexpected. This paper documents some of those challenges and presents the lessons learned so that other nuclear cleanup projects can benefit from the experience at Hanford.

  5. LESSONS LEARNED FROM CLEANING OUT THE SLUDGE FROM THE SPENT FUEL STORAGE BASINS AT HANFORD ICEM-07

    International Nuclear Information System (INIS)

    KNOLLMEYER PM

    2007-01-01

    Until 2004, the K Basins at Hanford, in southeastern Washington State, held the largest collection of spent nuclear fuel in the United States Department of Energy (DOE) complex. The K East and K West Basins are massive pools each holding more than 4 million liters of water - that sit less than 450 meters from the Columbia River. In a significant multi-year campaign that ended in 2004, Fluor Hanford removed all of the fuel from the two Basins, over 2,300 metric tons (4.6 million pounds), dried it, and then placed it into dry storage in a specially designed facility away from the River. Removing the fuel, however, did not finish the cleanup work at the K Basins. The years of underwater storage had corroded the metallic uranium fuel, leaving behind a thick and sometimes hard-packed layer of sludge that coated the walls, floors and equipment inside the Basins. In places, the depth of the sludge was measured in feet rather than inches, and its composition was definitely not uniform. Together the Basins held an estimated 50 cubic meters of sludge (42 cubic meters in K East and 8 cubic meters in K West). The K East sludge retrieval and transfer work was completed in May 2007. Vacuuming up the sludge into large underwater containers in each of the Basins and then consolidating it all in containers in the K West Basin have presented significant challenges, some unexpected. This paper documents some of those challenges and presents the lessons learned so that other nuclear cleanup projects can benefit from the experience at Hanford

  6. Sampling and analysis plan for sludge located in fuel storage canisters of the 105-K West basin

    International Nuclear Information System (INIS)

    Baker, R.B.

    1997-01-01

    This Sampling and Analysis Plan (SAP) provides direction for the first sampling of sludge from the K West Basin spent fuel canisters. The specially developed sampling equipment removes representative samples of sludge while maintaining the radioactive sample underwater in the basin pool (equipment is described in WHC-SD-SNF-SDD-004). Included are the basic background logic for sample selection, the overall laboratory analyses required and the laboratory reporting required. These are based on requirements put forth in the data quality objectives (WHC-SD-SNF-DQO-012) established for this sampling and characterization activity

  7. Technology Development And Deployment Of Systems For The Retrieval And Processing Of Remote-Handled Sludge From Hanford K-West Fuel Storage Basin

    International Nuclear Information System (INIS)

    Raymond, R.E.

    2011-01-01

    In 2011, significant progress was made in developing and deploying technologies to remove, transport, and interim store remote-handled sludge from the 105-K West Fuel Storage Basin on the Hanford Site in south-central Washington State. The sludge in the 105-K West Basin is an accumulation of degraded spent nuclear fuel and other debris that collected during long-term underwater storage of the spent fuel. In 2010, an innovative, remotely operated retrieval system was used to successfully retrieve over 99.7% of the radioactive sludge from 10 submerged temporary storage containers in the K West Basin. In 2011, a full-scale prototype facility was completed for use in technology development, design qualification testing, and operator training on systems used to retrieve, transport, and store highly radioactive K Basin sludge. In this facility, three separate systems for characterizing, retrieving, pretreating, and processing remote-handled sludge were developed. Two of these systems were successfully deployed in 2011. One of these systems was used to pretreat knockout pot sludge as part of the 105-K West Basin cleanup. Knockout pot sludge contains pieces of degraded uranium fuel ranging in size from 600 μm to 6350 μm mixed with pieces of inert material, such as aluminum wire and graphite, in the same size range. The 2011 pretreatment campaign successfully removed most of the inert material from the sludge stream and significantly reduced the remaining volume of knockout pot product material. Removing the inert material significantly minimized the waste stream and reduced costs by reducing the number of transportation and storage containers. Removing the inert material also improved worker safety by reducing the number of remote-handled shipments. Also in 2011, technology development and final design were completed on the system to remove knockout pot material from the basin and transport the material to an onsite facility for interim storage. This system is scheduled

  8. Successful Deployment of System for the Storage and Retrieval of Spent/Used Nuclear Fuel from Hanford K-West Fuel Storage Basin-13051

    International Nuclear Information System (INIS)

    Quintero, Roger; Smith, Sahid; Blackford, Leonard Ty; Johnson, Mike W.; Raymond, Richard; Sullivan, Neal; Sloughter, Jim

    2013-01-01

    In 2012, a system was deployed to remove, transport, and interim store chemically reactive and highly radioactive sludge material from the Hanford Site's 105-K West Fuel Storage Basin that will be managed as spent/used nuclear fuel. The Knockout Pot (KOP) sludge in the 105-K West Basin was a legacy issue resulting from the spent nuclear fuel (SNF) washing process applied to 2200 metric tons of highly degraded fuel elements following long-term underwater storage. The washing process removed uranium metal and other non-uranium constituents that could pass through a screen with 0.25-inch openings; larger pieces are, by definition, SNF or fuel scrap. When originally retrieved, KOP sludge contained pieces of degraded uranium fuel ranging from 600 microns (μm) to 6350 μm mixed with inert material such as aluminum hydroxide, aluminum wire, and graphite in the same size range. In 2011, a system was developed, tested, successfully deployed and operated to pre-treat KOP sludge as part of 105-K West Basin cleanup. The pretreatment process successfully removed the vast majority of inert material from the KOP sludge stream and reduced the remaining volume of material by approximately 65 percent, down to approximately 50 liters of material requiring management as used fuel. The removal of inert material resulted in significant waste minimization and project cost savings because of the reduced number of transportation/storage containers and improvement in worker safety. The improvement in worker safety is a result of shorter operating times and reduced number of remote handled shipments to the site fuel storage facility. Additionally in 2011, technology development, final design, and cold testing was completed on the system to be used in processing and packaging the remaining KOP material for removal from the basin in much the same manner spent fuel was removed. This system was deployed and successfully operated from June through September 2012, to remove and package the last

  9. Fuel storage tank

    International Nuclear Information System (INIS)

    Peehs, M.; Stehle, H.; Weidinger, H.

    1979-01-01

    The stationary fuel storage tank is immersed below the water level in the spent fuel storage pool. In it there is placed a fuel assembly within a cage. Moreover, the storage tank has got a water filling and a gas buffer. The water in the storage tank is connected with the pool water by means of a filter, a surge tank and a water purification facility, temperature and pressure monitoring being performed. In the buffer compartment there are arranged catalysts a glow plugs for recombination of radiolysis products into water. The supply of water into the storage tank is performed through the gas buffer compartment. (DG) [de

  10. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  11. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  12. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  13. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  14. Compact nuclear fuel storage

    International Nuclear Information System (INIS)

    Kiselev, V.V.; Churakov, Yu.A.; Danchenko, Yu.V.; Bylkin, B.K.; Tsvetkov, S.V.

    1983-01-01

    Different constructions of racks for compact storage of spent fuel assemblies (FA) in ''coolin''g pools (CP) of NPPs with the BWR and PWR type reactors are described. Problems concerning nuclear and radiation safety and provision of necessary thermal conditions arising in such rack design are discussed. It is concluded that the problem of prolonged fuel storage at NPPs became Very actual for many countries because of retapdation of the rates of fuel reprocessing centers building. Application of compact storage racks is a promising solution of the problem of intermediate FA storage at NPPs. Such racks of stainless boron steel and with neutron absorbers in the from of boron carbide panels enable to increase the capacity of the present CP 2-2.6 times, and the period of FA storage in them up to 5-10 years

  15. Spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.

    1982-06-01

    Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR, PWR) within a given utility. In all cases, a full core discharge capability (full core reserve or FCR) is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. For the current AR capacity case the indicated storage requirements in the year 2000 are indicated to be 18,190 MTU; for the maximum capacity with transshipment case they are 11,320 MTU

  16. Spent fuel storage facility, Kalpakkam

    International Nuclear Information System (INIS)

    Shreekumar, B.; Anthony, S.

    2017-01-01

    Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements

  17. Corrosion surveillance in spent fuel storage pools

    International Nuclear Information System (INIS)

    Howell, J.P.

    1996-01-01

    In mid-1991, corrosion of aluminum-clad spent nuclear fuel was observed in the light-water filled basins at the Savannah River site. A corrosion surveillance program was initiated in the P, K, L-Reactor basins and in the Receiving Basin for Offsite Fuels (RBOF). This program verified the aggressive nature of the pitting corrosion and provided recommendations for changes in basin operations to permit extended longer term interim storage. The changes were implemented during 1994--1996 and have resulted in significantly improved basin water quality with conductivity in the 1--3 microS/cm range. Under these improved conditions, no new pitting has been observed over the last three years. This paper describes the corrosion surveillance program at SRS and what has been learned about the corrosion of aluminum-clad in spent fuel storage pools

  18. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Kurokawa, Hideaki; Kumagaya, Naomi; Oda, Masashi; Matsuda, Masami; Maruyama, Hiromi; Yamanaka, Tsuneyasu.

    1997-01-01

    The structure of a spent fuel storage rack is determined by the material, thickness, size of square cylindrical tubes (the gap between spent fuel assemblies and the square cylindrical tubes) and pitch of the arrangement (the gap between each of the square cylindrical tubes). In the present invention, the thickness and the pitch of the arrangement of the square tubes are optimized while evaluating subcriticality. Namely, when the sum of the thickness of the water gap at the outer side (the pitch of arrangement of the cylindrical tubes) and the thickness of the cylindrical tubes is made constant, the storage rack is formed by determining the thickness of the cylindrical tubes which is smaller than the optimum value among the combination of the thickness of the water gap at the outer side and that of the cylindrical tube under the effective multiplication factor to be performed. Then, the weight of the rack can be reduced, and the burden of the load on the bottom of the pool can be reduced. Further, the amount of the constitutional materials of the rack itself can be reduced thereby capable of reducing the cost for the materials of the rack. (T.M.)

  19. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  20. WWER spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  1. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

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

  2. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  3. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

    Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

    1979-01-01

    Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

  4. Dry spent fuel storage licensing

    International Nuclear Information System (INIS)

    Sturz, F.C.

    1995-01-01

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

  5. Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; the 118-H-6:6 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

    International Nuclear Information System (INIS)

    Appel, M.J.

    2006-01-01

    This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31)

  6. Economics of spent LWR fuel storage

    International Nuclear Information System (INIS)

    Clark, H.J.

    1980-01-01

    A low cost option for spent fuel inventories would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) storage facilities, and away-from-reactor (AFR) storage facilities. Fuel storage requirements will be met best by transfer of fuel or by re-racking existing reactor basins whenever these options are available. These alternatives represent not only the lowest cost storage options but also the most timely. Fuel can be shipped to other storage pools for about $10/kg depending on the distance, while costs for reracking range from $18 to 25/kg depending on the approach. These alternatives are recognized to face environmental and regulatory obstacles. However, such obstacles should be less severe than similar issues that would be encountered with AR or AFR basin storage. When storage requirements cannot be met by the first two options, the next least costly alternative for most utilities will be use of a Federal AFR. Storage cost of about $137/kg at an AFR are less costly than charges of up to $350/kg that could be incurred by the use of AR basins. AR basins are practical only when a utility requires storage capacity to accommodate annual additions of 100 MT or more of spent fuel. The large reactor complexes discharging this much feul are not currently those that require relief from fuel storage problems. A recent development in Germany may offer an AR alternative of dry storage in transportation/storage casks at a cost of $200/kg; however, this method has not yet been accepted and licensed for use in the US

  7. Compressed gas fuel storage system

    Science.gov (United States)

    Wozniak, John J.; Tiller, Dale B.; Wienhold, Paul D.; Hildebrand, Richard J.

    2001-01-01

    A compressed gas vehicle fuel storage system comprised of a plurality of compressed gas pressure cells supported by shock-absorbing foam positioned within a shape-conforming container. The container is dimensioned relative to the compressed gas pressure cells whereby a radial air gap surrounds each compressed gas pressure cell. The radial air gap allows pressure-induced expansion of the pressure cells without resulting in the application of pressure to adjacent pressure cells or physical pressure to the container. The pressure cells are interconnected by a gas control assembly including a thermally activated pressure relief device, a manual safety shut-off valve, and means for connecting the fuel storage system to a vehicle power source and a refueling adapter. The gas control assembly is enclosed by a protective cover attached to the container. The system is attached to the vehicle with straps to enable the chassis to deform as intended in a high-speed collision.

  8. Cesium removal from the fuel storage water at the Savannah River Site R-Building Disassembly Basin using 3M Empore reg-sign-membrane filter technology

    International Nuclear Information System (INIS)

    Oji, L.N.; Thompson, M.C.; Peterson, K.; May, C.; Kafka, T.M.

    1998-01-01

    This report describes results from a seven-day demonstration of the use of 3M Empore membrane filter loaded with ion exchange material (potassium cobalt hexacynoferrate (CoHex)) for cesium uptake from the R-Disassembly Basin at the Savannah River Site. The goal of the demonstration was to evaluate the ability of the Process Absorber Development unit (PADU), a water pre-filtration /CoHex configuration on a skid, to remove cesium from R-Disassembly Basin at a linear processing flow rate of 22.7 liters per minute (1,195.8 liters/minute/m 2 or 29.35 gallons/minute/ft 2 ). Over 210,000 liters (> 55,500 gallons) of R-Disassembly Basin water was processed through the PADU without a cesium breakthrough, that is, the effluent after treatment with CoHex, contained less than detectable amounts of radioactive cesium. Some of the observed advantages of the Empore membrane filter technology over conventional packed column ion exchange systems includes rapid flow rates without channeling effects, low volume secondary waste and fast extraction or rapid kinetics per unit of flow

  9. High density fuel storage rack

    International Nuclear Information System (INIS)

    Zezza, L.J.

    1980-01-01

    High storage density for spent nuclear fuel assemblies in a pool achieved by positioning fuel storage cells of high thermal neutron absorption materials in an upright configuration in a rack. The rack holds the cells at required pitch. Each cell carries an internal fuel assembly support, and most cells are vertically movable in the rack so that they rest on the pool bottom. Pool water circulation through the cells and around the fuel assemblies is permitted by circulation openings at the top and bottom of the cells above and below the fuel assemblies

  10. Spent fuel storage criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

    1995-10-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs.

  11. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    Amin, E.M.; Elmessiry, A.M.

    1995-01-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

  12. Spent LWR fuel-storage costs

    International Nuclear Information System (INIS)

    Clark, H.J.

    1981-01-01

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

  13. Spent fuel storage requirements, 1988

    International Nuclear Information System (INIS)

    1988-10-01

    Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs

  14. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

    This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.

  15. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  16. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

    Bodette, D.E.; Naegeli, R.E.

    1997-10-01

    This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

  17. Advanced compressed hydrogen fuel storage systems

    International Nuclear Information System (INIS)

    Jeary, B.

    2000-01-01

    Dynetek was established in 1991 by a group of private investors, and since that time efforts have been focused on designing, improving, manufacturing and marketing advanced compressed fuel storage systems. The primary market for Dynetek fuel systems has been Natural Gas, however as the automotive industry investigates the possibility of using hydrogen as the fuel source solution in Alternative Energy Vehicles, there is a growing demand for hydrogen storage on -board. Dynetek is striving to meet the needs of the industry, by working towards developing a fuel storage system that will be efficient, economical, lightweight and eventually capable of storing enough hydrogen to match the driving range of the current gasoline fueled vehicles

  18. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

    Dyck, H.P.

    1999-01-01

    The paper gives an overview of the meetings held on the technology and safety aspects of regional spent fuel storage facilities. The questions of technique, economy and key public and political issues will be covered as well as the aspects to be considered for implementation of a regional facility. (author)

  19. A central spent fuel storage in Sweden

    International Nuclear Information System (INIS)

    Gustafsson, B.; Hagberth, R.

    1978-01-01

    A planned central spent fuel storage facility in Sweden is described. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed. (author)

  20. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs

  1. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  2. Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

    International Nuclear Information System (INIS)

    1985-01-01

    As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr

  3. Horizontal modular dry irradiated fuel storage system

    Science.gov (United States)

    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).

  4. The prospects for dry fuel storage

    International Nuclear Information System (INIS)

    Harris, G.G.; Elliott, D.

    1994-01-01

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

  5. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes the general safety requirements applicable to the design of both wet and dry spent fuel storage facilities; Section 3 deals with the design requirements specific to either wet or dry storage. Recommendations for the auxiliary systems of any storage facility are contained in Section 4; these are necessary to ensure the safety of the system and its safe operation. Section 5 provides recommendations for establishing the quality assurance system for a storage facility. Section 6 discusses the requirements for inspection and maintenance that must be considered during the design. Finally, Section 7 provides guidance on design features to be considered to facilitate eventual decommissioning. 18 refs

  6. Special equipment support the fuel storage

    International Nuclear Information System (INIS)

    Vega, M. E.

    2014-01-01

    In the current juncture one of the keys to any company that works in a market that is as demanding as the nuclear, is its ability to developed new technological products that they can adapt to the different special situations/needs of nuclear Power Plants during their operating life. As an example, below are some of the specialized equipment that ENSA has been developing for more than thirty years that has been doing work in the area of fuel storage. (Author)

  7. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  8. Spent fuel storage for ISER plant

    International Nuclear Information System (INIS)

    Nakajima, Takasuke; Kimura, Yuzi

    1987-01-01

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

  9. Commentary on spent fuel storage at Morris operation

    International Nuclear Information System (INIS)

    Eger, K.J.; Zima, G.E.

    1979-10-01

    The General Electric Company is providing technical support to Battelle Pacific Northwest Laboratories in the analysis of the design, operation, and maintenance experience in the handling of nuclear fuel at the Independent Spent Fuel Storage Facility. The purpose of this report is to provide a description of spent fuel handling activities and systems, and an analysis of the storage performance as developed over the seven year operational history of the Morris Operation. Design considerations and performance are analyzed for both the basin and key supporting systems. The bases for this analysis are the provisions for containing radioactive by-product materials, for shielding from the radiation they emit, and for preventing the formation of a critical array. These provisions have been met effectively over the history of storage at Morris. The release of radioactive materials is minimized by the protection of the cladding integrity, the containment of the basin water, the removal of radioactive and other contaminants from the water, and by filtering and then dispersing the basin air. Four auxiliary systems are provided to accomplish this, the basin leak detection system, the filter, the coolers, and the building ventilation system. This successful history notwithstanding, action to reduce personnel exposure, to improve fuel handling reliability and to lessen the potential for accidents continues to be taken

  10. Interface magnons. Magnetic superstructure

    International Nuclear Information System (INIS)

    Djafari-Rouhani, B.; Dobrzynski, L.

    1975-01-01

    The localized magnons at an interface between two Heisenberg ferromagnets are studied with a simple model. The effect of the coupling at the interface on the existence condition for the localized modes, the dispersion laws and the possible occurrence of magnetic superstructures due to soft modes are investigated. Finally a comparison is made with the similar results obtained for interface phonons [fr

  11. Nanoparticle assemblies and superstructures

    National Research Council Canada - National Science Library

    Kotov, Nicholas A

    2006-01-01

    ... building blocks of larger and more complex systems. Therefore, the present challenge of nanoscale science is to shift from making certain building blocks to organizing them in one-, two-, and three-dimensional structures. Such assemblies and superstructures are the next logical step in the development of nanoscience and nanotechnology. In this re...

  12. Reracking to increase spent fuel storage capacity

    International Nuclear Information System (INIS)

    1980-05-01

    Many utilities have already increased their spent fuel pool storage capacity by replacing aluminum racks having storage densities as low as 0.2 MTU/ft 2 with stainless steel racks which can more than double storage densities. Use of boron-stainless steel racks or thin stainless steel cans containing reassembled fuel rods allows even higher fuel storage densities (up to approximately 1.25 MTU/ft 2 ). This report evaluates the economics of smaller storage gains that occur if pools, already converted to high density storage, are further reracked

  13. Laser surveillance systems for fuel storage pools

    International Nuclear Information System (INIS)

    Boeck, H.

    1985-06-01

    A Laser Surveillance System (LASSY) as a new safeguards device has been developed under the IAEA research contract No. 3458/RB at the Atominstitut Wien using earlier results by S. Fiarman. This system is designed to act as a sheet of light covering spent fuel assemblies in spent fuel storage pools. When movement of assemblies takes place, LASSY detects and locates the position of the movement in the pool and when interrogated, presents a list of pool positions and times of movement to the safeguards inspector. A complete prototype system was developed and built. Full scale tests showed the principal working capabilities of a LASSY underwater

  14. Spent Fuel Storage Operation - Lessons Learned

    International Nuclear Information System (INIS)

    2013-12-01

    Experience gained in planning, constructing, licensing, operating, managing and modifying spent fuel storage facilities in some Member States now exceeds 50 years. Continual improvement is only achieved through post-project review and ongoing evaluation of operations and processes. This publication is aimed at collating and sharing lessons learned. Hopefully, the information provided will assist Member States that already have a developed storage capability and also those considering development of a spent nuclear fuel storage capability in making informed decisions when managing their spent nuclear fuel. This publication is expected to complement the ongoing Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III); the scope of which prioritizes facility operational practices in lieu of fuel and structural components behaviour over extended durations. The origins of the current publication stem from a consultants meeting held on 10-12 December 2007 in Vienna, with three participants from the IAEA, Slovenia and USA, where an initial questionnaire on spent fuel storage was formulated (Annex I). The resultant questionnaire was circulated to participants of a technical meeting, Spent Fuel Storage Operations - Lessons Learned. The technical meeting was held in Vienna on 13-16 October 2008, and sixteen participants from ten countries attended. A consultants meeting took place on 18-20 May 2009 in Vienna, with five participants from the IAEA, Slovenia, UK and USA. The participants reviewed the completed questionnaires and produced an initial draft of this publication. A third consultants meeting took place on 9-11 March 2010, which six participants from Canada, Hungary, IAEA, Slovenia and the USA attended. The meeting formulated a second questionnaire (Annex II) as a mechanism for gaining further input for this publication. A final consultants meeting was arranged on 20-22 June 2011 in Vienna. Six participants from Hungary, IAEA, Japan

  15. Structure for nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Ebata, Sakae; Nichiei, Shinji.

    1979-01-01

    Purpose: To enable leak detection in nuclear fuel storage pools, as well as prevent external leakages while keeping the strength of the constructional structures. Constitution: Protection plates are provided around pool linear plates and a leak reception is provided to the bottom. Leakages are detected by leak detecting pipeways and the external leakages are prevented by collecting them in a detection area provided in the intermediate layer. Since ferro-reinforcements at the bottom wall of the pool are disconnected by the protection plate making it impossible to form the constructional body, body hunches are provided to the bottom wall of the pool for processing the ferro-reinforcements. (Yoshino, Y.)

  16. Spent-fuel storage: a private sector option

    International Nuclear Information System (INIS)

    Thomas, J.A.; Ross, S.R.

    1983-01-01

    The investigation was performed to delineate the legal and financial considerations for establishing private sector support for the planning and development of an independent spent-fuel storage facility (ISFSF). The preferred institutional structure was found to be one in which a not-for-profit corporation contracts with a limited partnership to handle the spent fuel. The limited partnership acquires the necessary land and constructs the ISFSF facility and then leases the facility to the not-for-profit corporation, which acquires spent-fuel rods from the utilities. The DOE must agree to purchase the spent-fuel rods at the expiration of term and warrant continued operation of the facility if policy changes at the federal level force the removal of the rods prior to completion of the contracted storage cycle. The DOE planning base estimate of spent-fuel storage requirements indicates a market potential adequate to support 10,000 MTU or more of spent-fuel storage prior to the time a government repository is available to accept spent fuel around the turn of the century. The estimated construction cost of a 5000-MTU water basin facility is $552 million. The total capital requirements to finance such a facility are estimated to be $695 million, based on an assumed capital structure of 70 percent debt and 30 percent equity. The estimated total levelized cost of storage, including operating costs, for the assumed 17-year life of the facility is $223 per kilogram of uranium. This is equivalent to a slightly less than one mill per kilowatt-hour increase in nuclear fuel costs at the nuclear power station that was the source of the spent fuel. In conclusion, within the context of the new Nuclear Waste Policy Act of 1982, the study points to both the need for and the advantages of private sector support for one or more ISFSFs and establishes a workable mechanism for the recovery of the costs of owning and operating such facilities. 3 figures, 4 tables

  17. Superstructure of stapes

    International Nuclear Information System (INIS)

    Horiuchi, Yasuharu; Hunai, Hiroaki; Ichimura; Keiiti; Iinuma, Toshitaka; Oyama, Kazuyuki

    1989-01-01

    High-resolution Computed Tomography (HRCT) of the temporal bone has now become a routine test for the diagnosis of various temporal bone lesions. Correct assessment of such minute structures as ossicles, especially stapes, is important in the pre-operative HRCT evaluation. On the other hand, analysis for the reliability of HRCT findings has not been done, including the superstructure of the stapes. A retrospectively study was done in order to assess the availability of HRCT findings in axial overlapping scans in 226 ears, with respect to the superstructures of stapes. The study was concerned with the analysis of HRCTs of 148 normal ears, 71 ears of chronic otitis media and 7 ears of ossicular abnormalities. HRCT findings were compared to those of surgeries in pathological cases. The present study revealed that the superstructures of stapes are noted in 70% when the stapes is surrounded by air. The major limitations in the proper analysis of stapes by HRCT are partial volume averaging and effects of soft tissue silhouetting, this is especially so when the stapes is surrounded by soft tissue density. (author)

  18. Microbial Condition of Water Samples from Foreign Fuel Storage Facilities

    International Nuclear Information System (INIS)

    Berry, C.J.

    1998-01-01

    In order to assess the microbial condition of foreign spent nuclear fuel storage facilities and their possible impact on SRS storage basins, twenty-three water samples were analyzed from 12 different countries. Fifteen of the water samples were analyzed and described in an earlier report (WSRC-TR-97-00365 [1]). This report describes nine additional samples received from October 1997 through March 1998. The samples include three from Australia, two from Denmark and Germany and one sample from Italy and Greece. Each water sample was analyzed for microbial content and activity as determined by total bacteria, viable aerobic bacteria, viable anaerobic bacteria, viable sulfate-reducing bacteria, viable acid-producing bacteria and enzyme diversity. The results for each water sample were then compared to all other foreign samples analyzed to date and monthly samples pulled from the receiving basin for off-site fuel (RBOF), at SRS. Of the nine samples analyzed, four samples from Italy, Germany and Greece had considerably higher microbiological activity than that historically found in the RBOF. This microbial activity included high levels of enzyme diversity and the presence of viable organisms that have been associated with microbial influenced corrosion in other environments. The three samples from Australia had microbial activities similar to that in the RBOF while the two samples from Denmark had lower levels of microbial activity. These results suggest that a significant number of the foreign storage facilities have water quality standards that allow microbial proliferation and survival

  19. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    1987-04-01

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  20. Spent fuel storage requirements 1989--2020

    International Nuclear Information System (INIS)

    1989-10-01

    Historical inventories of spent fuel are combined with Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the US to provide estimates of spent fuel storage requirements over the next 32 years, through the year 2020. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Historical data through December 1988 are derived from the 1989 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 14 refs., 3 figs., 28 tabs

  1. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    1994-01-01

    The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

  2. International safeguards for spent fuel storage

    International Nuclear Information System (INIS)

    Kratzer, M.; Wonder, E.; Immerman, W.; Crane, F.

    1981-08-01

    This report analyzes the nonproliferation effectiveness and political and economic acceptability of prospective improvements in international safeguard techniques for LWR spent fuel storage. Although the applicability of item accounting considerably eases the safeguarding of stored spent fuel, the problem of verification is potentially serious. A number of simple gamma and neutron nondestructive assay techniques were found to offer considerable improvements, of a qualitative rather than quantitative nature, in verification-related data and information, and possess the major advantage of intruding very little on facility operations. A number of improved seals and monitors appear feasible as well, but improvements in the timeliness of detection will not occur unless the frequency of inspection is increased or a remote monitoring capability is established. Limitations on IAEA Safeguards resources and on the integration of results from material accounting and containment and surveillance remain problems

  3. Interim dry fuel storage for magnox reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, N [National Nuclear Corporation, Risley, Warrington (United Kingdom); Ealing, C [GEC Energy Systems Ltd, Whetstone, Leicester (United Kingdom)

    1985-07-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility.

  4. Realistic evaluation of new fuel storage criticality

    International Nuclear Information System (INIS)

    Hamasaki, M.; Itahara, K.; Shimada, S.; Etsu, M.; Kuroda, M.; Watanabe, H.; Kuragasaki, M.; Kawaguchi, K.

    1987-01-01

    In the criticality safety design of dry fuel storage facility, the optimum moderation condition which appears at rather low water density (0.05 ∼ 0.1 gH 2 O/cc) has been a barrier to incorporate economical design. In this study we have shown that this optimum moderation does not occur, even in fire-fighting with water supply system, by criticality analyses and experiments. Evaluated critical amount of water in the space has found itself far from the level estimated actually attainable through the experiment with fire-fighting system. In another experiment we have taken the holograms of 0.003 ∼ 0.35 g/cc water which was realized in the very narrow space by the nozzle manufactured for this special purpose. Those holograms demonstrate that the droplets in the low density water are more closely packed than we anticipate they are. (author)

  5. Interim dry fuel storage for magnox reactors

    International Nuclear Information System (INIS)

    Bradley, N.; Ealing, C.

    1985-01-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility

  6. Spent fuel storage requirements, 1991--2040

    International Nuclear Information System (INIS)

    1991-12-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 50 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1990 are derived from the 1991 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges

  7. Spent fuel storage requirements 1993--2040

    International Nuclear Information System (INIS)

    1994-09-01

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges

  8. Spent fuel storage requirements, 1990--2040

    International Nuclear Information System (INIS)

    Walling, R.; Bierschbach, M.

    1990-11-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 51 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1989 are derived from the 1990 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 15 refs., 3 figs., 11 tabs

  9. Features and safety aspects of spent fuel storage facility, Tarapur

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    Spent Fuel Storage Facility (SFSF), Tarapur is designed to store spent fuel arising from PHWRs in different parts of the country. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Tarapur was hot commissioned after regulatory clearances

  10. Spent nuclear fuel storage - Basic concept

    International Nuclear Information System (INIS)

    Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

    2009-01-01

    According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)

  11. Spent fuel storage options: a critical appraisal

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

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

  13. Nuclear reactor spent fuel storage rack

    International Nuclear Information System (INIS)

    Machado, O.J.; Flynn, W.M.; Flanders, H.E. Jr.; Booker, L.W.

    1989-01-01

    A fuel rack is described for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, the base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each cell positioned in a corresponding pair of the aligned polygonal openings, each cell being open at both ends with a guiding funnel at the upper end, and the cells being positioned over the flow openings in the base-plate to permit flow of coolant through the cells; spaced, outwardly directed, projections on the outer surfaces of the sides of the cells near the tops and bottoms of the sides thereof, each cell being sized to be received within a corresponding of the pair of aligned polygonal openings in which the cells are respectively positioned; and means fixedly securing the projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design

  14. Compacted spent-fuel storage--designs and problems

    International Nuclear Information System (INIS)

    Rubinstein, H.J.; Gilcrest, J.D.; Kendall, W.R.

    1979-01-01

    Typical rack designs, licensing, contracting methods, installation and operational problems are described. Due to the lack of reprocessing and independent fuel storage facilities, new plants built in the United States will continue to install high-density spent-fuel storage racks. As to the rack designs, the most significant feature is the introduction of freestanding rack designs. The trends in spent-fuel storage appear to be toward the use of high-density racks, either with or without absorber, for all plants in the design, construction, or operation phase; the use of freestanding rack designs; and the separation of engineering and fabrication during procurement

  15. Analysis of the impact of retrievable spent fuel storage

    International Nuclear Information System (INIS)

    Merrill, E.T.; White, M.K.; Fleischman, R.M.

    1978-03-01

    The impact of retrievably storing spent fuel is measurable in terms of the contribution the stored spent fuel makes to implementing the fuel management option selected. For the case of a decision to recycle LWR fuel in LWRs, a useful indicator of impact is the ratio of energy production with varying degrees of spent fuel retrievability to that achievable with total spent fuel retrievability. For a decision made in the year 2000, this ratio varies from 0.81 (10 yr storage in reactor basins) to 0.97 (retrievable storage for 25 years after fuel discharge). An earlier decision to recycle in LWRs results in both of these ratios being nearer to 1.0. If a decision is reached to implement a breeder reactor economy, the chosen comparison is the installed breeder capacity achievable with varying degrees of spent fuel retrievability. If a decision to build breeder reactors is reached in the year 2000, the maximum possible installed breeder capacity in 2040 varies from 490 GWe (10 yr storage in reactor basins) to 660 GWe (all fuel retrievably stored). If all fuel is retrievably stored 25 years, 635 GWe of breeder capacity is achievable by 2040. For an earlier decision date, such as 1985, the maximum possible installed breeder capacity in 2040 ranges from 740 GWe (no retrievable storage) to 800 GWe (all fuel retrievably stored). As long as a decision to reprocess is reached before 2000, most of the potential benefit of retrievable storage may be realized by implementing retrievable storage after such a decision is made. Neither providing retrievable spent fuel storage prior to a decision to reprocess, nor designing such storage for more than 25 years of retrievability appear to offer significant incremental benefit

  16. Studies and research concerning BNFP: LWR spent fuel storage

    International Nuclear Information System (INIS)

    Shallo, F.A.

    1978-08-01

    This report describes potential spent fuel storage expansion programs using the Barnwell Nuclear Fuel Plant--Fuel Receiving and Storage Station (BNFP-FRSS) as a model. Three basic storage arrangements are evaluated with cost and schedule estimates being provided for each configuration. A general description of the existing facility is included with emphasis on the technical and equipment requirements which would be necessary to achieve increased spent fuel storage capacity at BNFP-FRSS

  17. Method of storing the fuel storage pot in a fuel storage tank for away-from-reactor-storage

    International Nuclear Information System (INIS)

    Ishiguro, Jun-ichi.

    1980-01-01

    Purpose: To prevent the contact of sodium in the away-from-reactor-storage fuel storage tank with sodium in a fuel storage pool having radioactivity ana always retain clean state therein. Method: Sodium is filled in a container body of the away-from-reactor-storage fuel storage tank, and a conduit, a cycling pump, and cooling means are disposed to form a sodium coolant cycling loop. The fuel storage pool is so stored in the container body that the heat of the pool is projected from the liquid surface of the sodium in the container. Therefore, the sodium in the container is isolated from the sodium in the pool containing strong radioactivity to prevent contact of the former sodium from the latter sodium. (Sekiya, K.)

  18. Alternatives for water basin spent fuel storage using pin storage

    International Nuclear Information System (INIS)

    Viebrock, J.M.; Carlson, R.W.

    1979-09-01

    The densest tolerable form for storing spent nuclear fuel is storage of only the fuel rods. This eliminates the space between the fuel rods and frees the hardware to be treated as non-fuel waste. The storage density can be as much as 1.07 MTU/ft 2 when racks are used that just satisfy the criticality and thermal limitations. One of the major advantages of pin storage is that it is compatible with existing racks; however, this reduces the storage density to 0.69 MTU/ft 2 . Even this is a substantial increase over the 0.39 MTU/ft 2 that is achievable with current high capacity stainless steel racks which have been selected as the bases for comparison. Disassembly requires extensive operation on the fuel assembly to remove the upper end fitting and to extract the fuel rods from the assembly skeleton. These operations will be performed with the aid of an elevator to raise the assembly where each fuel rod is grappled. Lowering the elevator will free the fuel rod for transfer to the storage canister. A storage savings of $1510 per MTU can be realized if the pin storage concept is incorporated at a new away-from-reactor facility. The storage cost ranges from $3340 to $7820 per MTU of fuel stored with the lower cost applying to storage at an existing away-from-reactor storage facility and the higher cost applying to at-reactor storage

  19. Economics of spent LWR fuel storage

    International Nuclear Information System (INIS)

    Clark, H.J.; O'Neill, G.F.

    1980-01-01

    A power reactor operator, confronted with rising spent fuel inventories that would soon exceed his storage capacity, has to decide what to do with this fuel if he wants to continue reactor operations. A low cost option would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) basins for storage, and away-from-reactor (AFR) basins for storage. Economic considerations for each of the alternatives are compared

  20. Comparison of concepts for independent spent fuel storage facilities

    International Nuclear Information System (INIS)

    Held, Ch.; Hintermayer, H.P.

    1978-01-01

    The design and the construction costs of independent spent fuel storage facilities show significant differences, reflecting the fuel receiving rate (during the lifetime of the power plant or within a very short period), the individual national policies and the design requirements in those countries. Major incremental construction expenditures for storage facilities originate from the capacity and the type of the facilities (casks or buildings), the method of fuel cooling (water or air), from the different design of buildings, the redundancy of equipment, an elaborate quality assurance program, and a single or multipurpose design (i.e. interim or long-term storage of spent fuel, interim storage of high level waste after fuel storage). The specific costs of different designs vary by a factor of 30 to 60 which might in the high case increase the nuclear generating costs remarkably. The paper also discusses the effect of spent fuel storage on fuel cycle alternatives with reprocessing or disposal of spent fuel. (author)

  1. Dry spent fuel storage in the 1990's

    International Nuclear Information System (INIS)

    Roberts, J.P.

    1991-01-01

    In the US, for the decade of the 1990's, at-reactor-site dry spent fuel storage has become the predominant option outside of reactor spent fuel pools. This development has resulted from failure, in the 1980's, of a viable reprocessing option for commercial power reactors, and delay in geologic repository development to an operational date at or beyond the year 2010. Concurrently, throughout the 1980's, aggressive technical and regulatory efforts by the Federal Government, coordinated with nuclear industry, have led to successful evolution of dry spent fuel storage as a utility option

  2. Behavior of high-density spent-fuel storage racks

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1986-08-01

    Included in this report is a summary of information on neutron-absorbing materials such as B 4 C in an aluminum matrix or organic binder material, stainless steel-boron and aluminum-boron alloys, and stainless steetl-clad cadmium that are used in high-density spent fuel storage racks. A list of the types of neutron-absorbing materials being used in spent fuel storage racks at domestic commercial plants is provided. Recent cases at several domestic plants where swelling of rack side plates (where the B 4 C in an aluminum matrix and B 4 C in an organic binder material were located) occurred are reviewed

  3. Decontamination of FAST (CPP-666) fuel storage area stainless steel fuel storage racks

    International Nuclear Information System (INIS)

    Kessinger, G.F.

    1993-10-01

    The purpose of this report was to identify and evaluate alternatives for the decontamination of the RSM stainless steel that will be removed from the Idaho Chemical Processing plant (ICPP) fuel storage area (FSA) located in the FAST (CPP-666) building, and to recommend decontamination alternatives for treating this material. Upon the completion of a literature search, the review of the pertinent literature, and based on the review of a variety of chemical, mechanical, and compound (both chemical and mechanical) decontamination techniques, the preliminary results of analyses of FSA critically barrier contaminants, and the data collected during the FSA Reracking project, it was concluded that decontamination and beneficial recycle of the FSA stainless steel produced is technically feasible and likely to be cost effective as compared to burying the material at the RWMC. It is recommended that an organic acid, or commercial product containing an organic acid, be used to decontaminate the FSA stainless steel; however, it is also recommended that other surface decontamination methods be tested in the event that this method proves unsuitable. Among the techniques that should be investigated are mechanical techniques (CO 2 pellet blasting and ultra-high pressure water blasting) and chemical techniques that are compatible with present ICPP waste streams

  4. A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL & TREATMENT & DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

    Energy Technology Data Exchange (ETDEWEB)

    PERES, M.W.

    2007-01-09

    Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel.

  5. Commercial solutions [for dry spent fuel storage casks

    International Nuclear Information System (INIS)

    Howe, W.F.; Pennington, C.W.; Hobbs, J.; Lee, W.; Thomas, B.D.; Dibert, D.J.

    1996-01-01

    In the aftermath of the termination of the DOE's MPC (Multi-Purpose Canister) programme, commercial suppliers are coming forward with new or updated systems to meet utility needs. Leading vendors describe the advantages of their systems for dry spent fuel storage and transport. (Author)

  6. How Canada has controlled the spent fuel storage problem

    International Nuclear Information System (INIS)

    Mosey, D.

    1985-01-01

    A report on the irradiated fuel storage workshop held in Toronto in October 1984. In particular Canada's attitude to spent fuel is examined. The basic fuel cycle has been envisaged as running from mining and refining, through interim storage to final geologic disposal, with reprocessing as an option to be considered when it looks economically attractive. (U.K.)

  7. Independent Spent Fuel Storage Installations (ISFSI). Annual report, FY 1978

    International Nuclear Information System (INIS)

    Zima, G.E.

    1979-03-01

    The prime objective of the subject program is the identification of technical aspects of the design, operation and maintenance of independent spent fuel storage installations which could contribute to technical bases for Regulations and Regulatory Guides issued by NRC for these facilities. Activities on the various tasks of the program for the FY 1978 period are discussed in this report

  8. Contributions to LWR spent fuel storage and transport

    International Nuclear Information System (INIS)

    The papers included in this document describe the aspects of spent LWR fuel storage and transport-behaviour of spent fuel during storage; use of compact storage packs; safety of storage; design of storage facilities AR and AFR; description of transport casks and transport procedures

  9. Bulk Fuel Storage and Delivery Systems Infrastructure Military Construction Requirements for Japan

    National Research Council Canada - National Science Library

    Padgett, Gary

    2000-01-01

    .... Specifically, this audit evaluated requirements for bulk fuel storage facilities at three locations in Japan. We also evaluated the management control program as it relates to the bulk fuel storage military construction requirements validation process.

  10. Corrosion of aluminum-clad alloys in wet spent fuel storage

    International Nuclear Information System (INIS)

    Howell, J.P.

    1995-09-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

  11. 78 FR 63375 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear...] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No..., Inc. Standardized NUHOMS[supreg] Cask System listing within the ``List of Approved Spent Fuel Storage...

  12. 75 FR 27463 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

    Science.gov (United States)

    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...

  13. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... Fuel Storage Casks: MAGNASTOR [supreg] System, Revision 2 AGENCY: Nuclear Regulatory Commission. ACTION... its spent fuel storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR [supreg] System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2 to...

  14. Existing and near future practices of spent fuel storage in Slovak Republic

    International Nuclear Information System (INIS)

    Mizov, J.

    1999-01-01

    In this paper existing and near future practices of spent fuel storage in Slovak Republic are discussed: (1) Reactor operation and spent fuel production; (2) Past policy in spent fuel storage; (3) Away-from-reactor (AFR) storage facility at Bohunice NPP site; (4) Present policy in spent fuel storage; (5) Final disposal of spent fuel

  15. Spent fuel storage at Prairie Island: January 1995 status

    International Nuclear Information System (INIS)

    Closs, J.; Kress, L.

    1995-01-01

    The disposal of spent nuclear fuel has been an issue for the US since the inception of the commercial nuclear power industry. In the past decade, it has become a critical factor in the continued operation of some nuclear power plants, including the two units at Prairie Island. As the struggles and litigation over storage alternatives wage on, spent fuel pools continue to fill and plants edge closer to premature shutdown. Due to the delays in the construction of a federal repository, many nuclear power plants have had to seek interim storage alternatives. In the case of Prairie Island, the safest and most feasible option is dry cask storage. This paper discusses the current status of the Independent Spent Fuel Storage Installation (ISFSI) Project at Prairie Island. It provides a historical background to the project, discusses the notable developments over the past year, and presents the projected plans of the Northern States Power Company (NSP) in regards to spent fuel storage

  16. Projection of US LWR spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

    1982-11-01

    The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

  17. The united kingdom's changing requirements for spent fuel storage

    International Nuclear Information System (INIS)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N.

    2013-01-01

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK

  18. Evaluation of economics of spent fuel storage techniques

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

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

  19. Thermal-hydraulic analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

  20. Standard plans for timber bridge superstructures

    Science.gov (United States)

    James P. Wacker; Matthew S. Smith

    2001-01-01

    These standardized bridge plans are for superstructures consisting of treated timber. Seven superstructure types are includes: five longitudinal and two transverse deck systems. Both HS520 and HS25 loadings are included, along with L/360 and L/500 deflection criteria.

  1. Implementing Composite Superstructures in Large Passenger Ships

    DEFF Research Database (Denmark)

    Karatzas, Vasileios; Berggreen, Christian; Jensen, Jørgen Juncher

    2015-01-01

    This study focuses on the structural response of the part of the superstructure of a RoPax ferry that has been redesigned using composite materials. The composite superstructure is presented and subsequently compared to the existing steel design considering different loading conditions by the use...

  2. Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

    International Nuclear Information System (INIS)

    Oh, Jinho

    2013-01-01

    The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe

  3. Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe.

  4. Stabilization of reactor fuel storage pool-TTP

    International Nuclear Information System (INIS)

    Sevigny, G.

    1994-10-01

    The proposed work includes evaluating standard and improved technologies an designing an integrated demonstration system to clean the water and sludge the fuel storage pools. The water released would meet drinking water standards and tritium standards. The volume of radioactive sludge would be reduced by partial separation of the sludge and radionuclides and eventual solidification of the hazardous and radioactive waste. The scope of the wo includes a survey of needs and applicable technologies, system engineering evaluation, conceptual design, detailed design, fabrication of the integrat demonstration system, and testing of the system. The survey task will locate potential specific customers within the DOE complex, and outside of the DOE complex throughout the United States, that be able to utilize the narrowly focused technology to stabilize/shutdown reactor fuel storage pools, responsible parties will be located and asked respond to a survey about their specific process requirements. Literature searches will be run through technical and scientific databases to locate technologies that may be an improvement over the standard baselined technol for cleanup of radioactively-contaminated pools. Systems engineering will provide decision analysis support for the development, evaluation, design, test functions of the treatment of pool water and sludge

  5. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  6. The Canadian long-term experimental used fuel storage program

    International Nuclear Information System (INIS)

    Wasywich, K.M.; Taylor, P.

    1993-01-01

    The Canadian experimental fuel storage program consists of four components: (1) storage of used CANDU (CANadian Deuterium Uranium, registered trademark of AECL) fuel under water, with periodic examination; (2) storage of used CANDU fuel in dry air at seasonally varying temperatures, and in both dry and moisture-saturated air at 150 C, also with periodic examination; (3) underlying research on the oxidation of unused and used UO 2 in dry and moist air at temperatures up to 300 C; and (4) modeling of UO 2 oxidation in dry air. The primary objective of the fuel-storage experiments is to investigate the stability of used CANDU fuel during long-term storage. Burnup of the fuel in these experiments ranges from ∼43 to 582 MW h/kg U, while the outer-element linear power ratings range from 22 to 79 kW/m. The storage behavior of intact and intentionally defected fuel, and fuel that defected in-reactor, is being investigated in the above experiments. Since differences in UO 2 oxidation behavior were observed between dry-air, moisture-saturated air and wet storage of intentionally defected used CANDU fuel, underlying research was initiated on oxidation of unused and used fuel to develop a better understanding of the different mechanisms. Modeling of UO 2 oxidation based on the results of the dry-storage experiments is also under way

  7. Stabilization of reactor fuel storage pool-TTP

    Energy Technology Data Exchange (ETDEWEB)

    Sevigny, G.

    1994-10-01

    The proposed work includes evaluating standard and improved technologies an designing an integrated demonstration system to clean the water and sludge the fuel storage pools. The water released would meet drinking water standards and tritium standards. The volume of radioactive sludge would be reduced by partial separation of the sludge and radionuclides and eventual solidification of the hazardous and radioactive waste. The scope of the wo includes a survey of needs and applicable technologies, system engineering evaluation, conceptual design, detailed design, fabrication of the integrat demonstration system, and testing of the system. The survey task will locate potential specific customers within the DOE complex, and outside of the DOE complex throughout the United States, that be able to utilize the narrowly focused technology to stabilize/shutdown reactor fuel storage pools, responsible parties will be located and asked respond to a survey about their specific process requirements. Literature searches will be run through technical and scientific databases to locate technologies that may be an improvement over the standard baselined technol for cleanup of radioactively-contaminated pools. Systems engineering will provide decision analysis support for the development, evaluation, design, test functions of the treatment of pool water and sludge.

  8. Benchmarking criticality analysis of TRIGA fuel storage racks.

    Science.gov (United States)

    Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

    2017-01-01

    A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

  9. Bulk Fuel Storage Requirements for Maintenance, Repair, and Environmental Projects at Fort Hood, Texas

    National Research Council Canada - National Science Library

    Carros, Deborah

    2000-01-01

    This report is one in a series that addresses the accuracy and reliability of maintenance, repair, environmental, and construction requirements for bulk fuel storage and delivery systems infrastructure...

  10. Manufacturing and Construction of Spent Fuel Storage Rack for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangjin; Jung, Kwangsub; Oh, Jinho; Lee, Jongmin [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The spent fuel storage rack consists of spent fuel storage racks and support frame. The spent fuel storage racks are installed in the support frame. A spent fuel storage rack consists of frame weldment and storage cell pipe assembly. Storage cell pipe assembly is mounted on the base plate of the frame weldment. The spent fuel storage rack is designed to withstand seismic load and other loads during earthquake. The structural integrity of the spent fuel storage rack is evaluated in accordance with ASME Section III, Subsection NF. Computer Code used for this analysis is ANSYS version 14.0.0. Dead load and seismic load is considered in load condition and hydrodynamic mass is included in the analysis. Design, manufacturing, and construction of the spent fuel storage rack are introduced. The spent fuel storage rack is for storage of spent fuel assemblies. The spent fuel storage rack should be designed, manufactured, and installed with consideration of predicted number of spent fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability.

  11. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    1999-07-01

    Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel

  12. Conceptual design report for the away from reactor spent fuel storage facility, Savannah River Plant

    International Nuclear Information System (INIS)

    1978-12-01

    The Department of Energy (DOE) requested that Du Pont prepare a conceptual design and appraisal of cost for Federal budget planning for an away from reactor spent fuel storage facility that could be ready to store fuel by December 1982. This report describes the basis of the appraisal of cost in the amount of $270,000,000 for all facilities. The proposed action is to provide a facility at the Savannah River Plant. The facility will have an initial storage capacity of 5000 metric tons of spent fuel and will be capable of receiving 1000 metric tons per year. The spent fuel will be stored in water-filled concrete basins that are lined with stainless steel. The modular construction of the facility will allow future expansion of the storage basins and auxiliary services in a cost-effective manner. The facility will be designed to receive, handle, decontaminate and reship spent fuel casks; to remove irradiated fuel from casks; to place the fuel in a storage basin; and to cool and control the quality of the water. The facility will also be designed to remove spent fuel from storage basins, load the spent fuel into shipping casks, decontaminated loaded casks and ship spent fuel. The facility requires a license by the Nuclear Regulatory Commission (NRC). Features of the design, construction and operations that may affect the health and safety of the workforce and the public will conform with NRC requirements. The facility would be ready to store fuel by January 1983, based on normal Du Pont design and construction practices for DOE. The schedule does not include the effect of licensing by the NRC. To maintain this option, preparation of the documents and investigation of a site at the Savannah River Plant, as required for licensing, were started in FY '78

  13. Superstructures: First Cold Test and Future Applications

    International Nuclear Information System (INIS)

    J. Sekutowicz; C Albrecht; V Ayvazyan; R Bandelmann; T Buttner; P Castro; S Choroba; J Eschke; B Faatz; A Gossel; K Honkavaara; B Horst; J Iverson; K Jensch; H Kaiser; R Kammering; G Kreps; D Kostin; J Lorkiewicz; R Lange; A Matheisen; W -D Moller; H -B Peters; D Proch; K Rehlich; H Schlarb; S Schrieber; D Reschke; S Simrock; W Singer; X Singer; K Twarowski; T Weichert; M Wojtkiewicz; G Wendt; K Zapfe; M Liepe; M Huening; M Ferrario; E Plawski; C Pagani; P Kneisel; G Wu; N Baboi; C Thomas; H Chen; W Huang; C Tang; S Zheng

    2003-01-01

    Superstructures, chains of superconducting multi-cell cavities (subunits) connected by e/2 long tube(s) have been proposed as an alternative layout for the TESLA main accelerator [1]. After three years of preparation, two superstructures, each made of two weakly coupled superconducting 7-cell subunits driven by a single Fundamental Power Coupler (FPC), have been installed in the Tesla Test Facility linac for beam tests. Energy stability, HOM damping, frequency and field adjustment methods were tested. The measured results confirme Superstructures, chains of superconducting multi-cell cavities (subunits) connected by e/2 long tube(s) have been proposed as an alternative layout for the TESLA main accelerator [1]. After three years of preparation, two superstructures, each made of two weakly coupled superconducting 7-cell subunits driven by a single Fundamental Power Coupler (FPC), have been installed in the Tesla Test Facility linac for beam tests. Energy stability, HOM damping, frequency and field adjustment methods were tested. The measured results confirmed expectation on the superstructure performance and proved that an alternative layout for the 800 GeV upgrade of the TESLA collider is feasible. We report on the test and give here an overview of its results. The tests confirmed very good damping of HOMs in superstructures and thus has opened a possible new application of this concept to high current energy recovery machines. We have built two 1.5 GHz copper models of two superstructures: 2x5-cells and 2x2-cells to prove further improvement of HOM damping. This contribution presents also measured results on these models. d expectations on the superstructure performance and proved that an alternative layout for the 800 GeV upgrade of the TESLA collider is feasible. We report on the test and give here an overview of its results

  14. Dry spent fuel storage facility at Kozloduy Nuclear Power Plant

    International Nuclear Information System (INIS)

    Goehring, R.; Stoev, M.; Davis, N.; Thomas, E.

    2004-01-01

    The Dry Spent Fuel Storage Facility (DSF) is financed by the Kozloduy International Decommissioning Support Fund (KIDSF) which is managed by European Bank for Reconstruction and Development (EBRD). On behalf of the Employer, the Kozloduy Nuclear Power Plant, a Project Management Unit (KPMU) under lead of British Nuclear Group is managing the contract with a Joint Venture Consortium under lead of RWE NUKEM mbH. The scope of the contract includes design, manufacturing and construction, testing and commissioning of the new storage facility for 2800 VVER-440 spent fuel assemblies at the KNPP site (turn-key contract). The storage technology will be cask storage of CONSTOR type, a steel-concrete-steel container. The licensing process complies with the national Bulgarian regulations and international rules. (authors)

  15. Life Prediction of Spent Fuel Storage Canister Material

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald

    2018-04-16

    The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.

  16. Economical evaluation on spent fuel storage technology away from reactor

    International Nuclear Information System (INIS)

    Itoh, Chihiro; Nagano, Koji; Saegusa, Toshiari

    2000-01-01

    Concerning the spent fuel storage away from reactor, economical comparison was carried out between metal cask and water pool storage technology. The economic index was defined by levelized cost (Unit storage cost) calculated on the assumption that the storage cost is paid at the receipt of the spent fuel at the storage facility. It is found that the cask storage is economical for small and large storage capacity. Unit storage cost of pool storage, however, is getting close to that of cask storage in case of storage capacity of 10,000 ton. Then, the unit storage cost is converted to power generation cost using data of the burn up of the fuel, etc. The cost is obtained as yen 0.09/kWh and yen 0. 15/kWh for cask storage and pool storage, respectively in case of the capacity of 5,000 tonU and the cooling time of 5 years. (author)

  17. Information handbook on independent spent fuel storage installations

    International Nuclear Information System (INIS)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996

  18. Method of assembling spent nuclear fuel storage rack

    International Nuclear Information System (INIS)

    Igarashi, Ryokichi; Hasegawa, Hidenobu.

    1982-01-01

    Purpose: To improve the safety of a spent fuel storage rack by stably installing the spent fuel in a pool without using supporting beams. Constitution: A restricted unit is composed of a plurality of spuare cylinders. A plurality of such restricted units are aligned in a direction perpendicularly to the arraying direction of the cylinders in the respective restricted units, are coupled with long connecting plates, and are fixed by welding on a common small base, thereby forming a restricted body. According to such assembling method, a plurality of restricted bodies are connected in a direction that the respective restricted bodies are readily overturned, and are secured to the common base. Accordingly, the restricted bodies can be stably installed in a pool without using supporting beams as the conventional one. (Sekiya, K.)

  19. Thermal stress analysis of the fuel storage facility

    International Nuclear Information System (INIS)

    Chen, W.W.

    1991-12-01

    This paper presents the results of a nonlinear finite-element analysis to determine the structural integrity of the walls of the nuclear fuel storage room in the Radio Isotope Power System Facility of the Fuels and Materials Examination Facility (FMEF) Project. The analysis was performed to assess the effects of thermal loading on the walls that would result from a loss-of-cooling accident. The results obtained from using the same three-dimensional finite-element model with different types of elements, the eight-node brick element and the nonlinear concrete element, and the calculated results using the analytical solutions, are compared. The concrete responses in terms of octahedral normal and shearing stresses are described. The crack and crush states of the concrete were determined on the basis of multiaxial failure criteria

  20. Remote inspection of the IFSF spent fuel storage rack

    International Nuclear Information System (INIS)

    Uldrich, E.D.

    1996-01-01

    The Irradiated Fuel Storage Facility (IFSF) is a dry storage facility for spent nuclear fuels located at the Idaho Chemical Processing Plant; it was constructed in the 1970's specifically for the Fort Saint Vrain spent reactor fuels. Currently, it is being used for various spent fuels. It was not known if IFSF would met current DOE seismic criteria, so re-analysis was started, with the rack being analyzed first. The rack was inspected to determine the as-built condition. LazrLyne and VideoRuler were used in lieu of using a tape measure with the camera. It was concluded that when a visual inspection shows widely varying weld sizes, the engineer has to use all resources available to determine the most probable specified weld sizes

  1. Spent fuel storage rack for atomic power plant

    International Nuclear Information System (INIS)

    Kodama, Tatemitsu.

    1981-01-01

    Purpose: To flexibly cope with the changes in the size and shape of spent fuel storage containers by placing a number of independently-constructed rack cells in a rack frame in such a manner that the guide support members of the storage rack, mounted on each rack cell may be replaced. Constitution: Independently-constructed rack cells are inserted from above into a rack frame rigidly installed on the bottom of a water pool. Each cell is produced by welding, has a handling head mounted at the top, and guide support members made of three replaceable guide tubes are mounted with bolts. If the size and the shape of the containers are altered, this configuration can easily cope with the new container shape by merely having the guide tubes replaced, without adversely affecting other cells and without necessitating draining of the water in the pool. (Yoshino, Y.)

  2. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  3. A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL, TREATMENT and DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

    International Nuclear Information System (INIS)

    PERES, M.W.

    2007-01-01

    Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel

  4. Status of spent fuel storage facilities in Switzerland

    International Nuclear Information System (INIS)

    Beyeler, P.C.; Lutz, H.R.; Heesen, W. von

    1999-01-01

    Planning of a dry spent fuel storage facility in Switzerland started already 15 years ago. The first site considered for a central interim storage facility was the cavern of the decommissioned pilot nuclear plant at Lucens in the French-speaking part of Switzerland. This project was terminated in the late eighties because of lack of public acceptance. The necessary acceptance was found in the small town of Wuerenlingen which has hosted for many years the Swiss Reactor Research Centre. The new project consists of centralised interim storage facilities for all types of radioactive waste plus a hot cell and a conditioning and incinerating facility. It represents a so-called integrated storage solution. In 1990, the new company 'ZWILAG Zwischenlager Wuerenlingen AG' (ZWILAG) was founded and the licensing procedures according to the Swiss Atomic law were initiated. On August 26, 1996 ZWILAG got the permit for construction of the whole facility including the operating permit for the storage facilities. End of construction and commissioning are scheduled for autumn 1999. The nuclear power station Beznau started planning a low level waste and spent fuel storage facility on its own, because in 1990 its management thought that by 1997 the first high active waste from the reprocessing facilities in France would have to be taken back. This facility at the Beznau site, called ZWIBEZ, was licensed according to a shorter procedure so its construction was finished by 1997. The two facilities for high level waste and spent fuel provide space for a total of 278 casks, which is sufficient for the waste and spent fuel of the four Swiss nuclear power stations including their life extension programme. (author)

  5. Risk assessment in spent fuel storage and transportation

    International Nuclear Information System (INIS)

    Pandimani, S.

    1989-01-01

    Risk assessment in various stages of nuclear fuel cycle is still an active area of Nuclear safety studies. From the results of risk assessment available in literature, it can be determined that the risk resulting from shipments of plutonium and spent-fuel are much greater than that resulting from the transport of other materials within the nuclear fuel cycle. In India spent fuels are kept in Spent Fuel Storage Pool (SFSP) for about 240-400 days, which is relatively a longer period compared to the usual 120 days as recommended by regulatory authorities. After cooling spent fuels are transported to the reprocessing sites which are mostly situated close to the plants. India has two high level waste treatment facilities, one PREFRE (Plutonium Reprocessing and Fuel Recycling) at Tarapur and the other one, a unit of Nuclear Fuel Complex at Hyderabad. This paper presents the risk associated with spent fuel storage and transportation for the Indian conditions. All calculations are based on a typical CANDU reactor system. Simple fault tree models are evolved for SFSP and for Transportation Accident Mode (TAM) for both road and rail. Fault tree quantification and risk assessment are done to each of these models. All necessary data for SFSP are taken mostly from Reactor Safety Study, (1975). Similarly, the data for rail TAM are taken from Annual Statistical Statements, (1987-8) and that for road TAM from Special Issue on Motor Vehicle Accident Statistics in India, (1986). Simulation method is used wherever necessary. Risk is also estimated for normal/accident free transport

  6. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    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...

  7. 76 FR 2243 - List of Approved Spent Fuel Storage Casks: NUHOMS ® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... Storage Casks: NUHOMS [supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION: Direct... fuel storage regulations by revising the Transnuclear, Inc. (TN) NUHOMS [supreg] HD System listing... NUHOMS [supreg] HD System cask design listed in Sec. 72.214 (List of approved spent fuel storage casks...

  8. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... Fuel Storage Casks: MAGNASTOR System, Revision 1 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) MAGNASTOR System listing within the ``List of...

  9. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel Storage Casks'' to include... for the MAGNASTOR[supreg] System cask design within the list of approved spent fuel storage casks that...

  10. 75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-06

    ... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... storage regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the... System cask design within the list of approved spent fuel storage casks that power reactor licensees can...

  11. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2010-05-07

    ...] RIN 3150-AI75 List of Approved Spent Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY...), NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include... Modular Storage System for Irradiated Nuclear Fuel. Docket Number: 72-1030. Certificate Expiration Date...

  12. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... Fuel Storage Casks: MAGNASTOR System, Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... storage cask regulations by revising the NAC International, Inc. (NAC), MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to Certificate of...

  13. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ...-2010-0183] RIN 3150--AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6 AGENCY.... (NAC), NAC-MPC System listing within the ``List of Approved Spent Fuel Storage Casks'' to include... changes to the configuration of the NAC-MPC storage system as noted in Appendix B of the Technical...

  14. 78 FR 32077 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-05-29

    ... Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct... All-purpose Storage (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel... CoC No. 1031, MAGNASTOR[supreg] System listing within the ``List of Approved Spent Fuel Storage Casks...

  15. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear...] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No... Safety Analysis Report for the Standardized NUHOMS[supreg] Horizontal Modular Storage System for...

  16. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Science.gov (United States)

    2011-01-13

    ... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to... the NUHOMS[supreg] HD Horizontal Modular Storage System for Irradiated Nuclear Fuel. [[Page 2279...

  17. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    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...

  18. Optimization of time and location dependent spent nuclear fuel storage capacity

    International Nuclear Information System (INIS)

    Macek, V.

    1977-01-01

    A linear spent fuel storage model is developed to identify cost-effective spent nuclear fuel storage strategies. The purpose of this model is to provide guidelines for the implementation of the optimal time-dependent spent fuel storage capacity expansion in view of the current economic and regulatory environment which has resulted in phase-out of the closed nuclear fuel cycle. Management alternatives of the spent fuel storage backlog, which is created by mismatch between spent fuel generation rate and spent fuel disposition capability, are represented by aggregate decision variables which describe the time dependent on-reactor-site and off-site spent fuel storage capacity additions, and the amount of spent fuel transferred to off-site storage facilities. Principal constraints of the model assure determination of cost optimal spent fuel storage expansion strategies, while spent fuel storage requirements are met at all times. A detailed physical and economic analysis of the essential components of the spent fuel storage problem, which precedes the model development, assures its realism. The effects of technological limitations on the on-site spent fuel storage expansion and timing of reinitiation of the spent fuel reprocessing on optimal spent fuel storage capacity expansion are investigated. The principal results of the study indicate that (a) expansion of storage capacity beyond that of currently planned facilities is necessary, and (b) economics of the post-reactor fuel cycle is extremely sensitive to the timing of reinitiation of spent fuel reprocessing. Postponement of reprocessing beyond mid-1982 may result in net negative economic liability of the back end of the nuclear fuel cycle

  19. 78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    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...

  20. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    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 until 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

  1. NRC analysis of the environmental impacts and licensing policies for expanded spent fuel storage in the United States

    International Nuclear Information System (INIS)

    Smith, C.V. Jr.

    1978-01-01

    This paper reviews the findings of the U.S. environmental impact studies on spent fuel storage to the year 2000, addresses design criteria for independent spent fuel storage installations and expresses the position of the NRC for licensing expanded spent fuel storage capabilities until such time as final solutions are established

  2. NRC Analysis of the environmental impacts and licensing policies for expanded spent fuel storage in the United States

    International Nuclear Information System (INIS)

    Clifford, V.S.

    1978-01-01

    The findings of the U.S. environmental impact studies on spent fuel storage to the year 2000 are reviewed, the design criteria for independent spent fuel storage installations are addressed and the position of the NRC for licensing expanded spent fuel storage capabilities, until such time as final solutions are established, is expressed. (author)

  3. Past experience and future needs for the use of burnup credit in LWR fuel storage

    International Nuclear Information System (INIS)

    Boyd, W.A.; Wrights, G.N.

    1987-01-01

    To achieve improved fuel economics and reduce the amount of fuel discharged annually, utilities are engaging in fuel management strategies that will achieve higher discharge burnups for their fuel assemblies. Although burnup credit methodologies have been developed and spent-fuel racks have been licensed, burnup credit fuel storage racks are not the answer for all utilities. Off-site and out-of-pool spent-fuel storage may be more appropriate. This is leading to the development of dry spent-fuel storage and shipping casks. Cask designs with spent-fuel storage capability between 20 and 32 assemblies are being developed by several vendors. The US Dept. of Energy is also funding work by VEPCO. Westinghouse is currently licensing its dry storage cask, developing a shipping cask for the domestic market, and is involved in a joint venture to develop a cask for the international market. Although methods of taking credit for fuel burnup in spent-fuel storage racks have been developed and licensed, use of these methods on dry spent-fuel storage and shipping casks can lead to new issues. These issues arise because the excess reactivity margin that is inherent in a burnup credit spent-fuel storage rack criticality analysis will not be available in a dry cask analysis

  4. Thermal Analysis Evaluation of Spent Fuel Storage Rack for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangjin; Oh, Jinho; Kwak, Jinsung; Lee, Jongmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Spent fuel storage rack is to store spent fuel assemblies. The spent fuel storage rack is submerged in the designated pool for cooling. Due to the condition change of the pool water, the effect of thermal load on spent fuel storage rack must be analyzed and evaluated. In this paper, thermal stress analysis is performed and evaluated on a spent fuel storage rack. For thermal stress evaluation of the spent fuel storage rack, load combinations and allowable criteria in ASME Sec. III NB-3220 are applied. In cases of A-1 and B-1, the same temperature applied on the whole model, thermal stress doesn't occur because there is no constraint about the thermal expansion. The support frame is located on the pool bottom in free standing type and the racks are located in the support frame with enough space. Thermal expansion was considered and reflected in the design of spent fuel storage rack in advance. Thermal stress analysis is performed and evaluated on a spent fuel storage rack with consideration of pool water temperature variation. The thermal analysis including a linear heat transfer and the thermal stress analysis is performed for the racks and support frame and resulted stresses are within allowable criteria.

  5. Estimated risk contribution for dry spent fuel storage cask

    International Nuclear Information System (INIS)

    Santos, C.; Kirk, M.T.; Abramson, L.; Guttmann, J.; Hackett, E.; Simonen, F.A.

    2001-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is pursuing means to risk-inform its regulations and programs for dry storage of spent nuclear fuel. In pursuit of this objective, the NRC will develop safety goals and probabilistic risk assessments for implementing risk-informed programs. This paper provides one example method for calculating the risk of a dry spent fuel storage cask under normal and accident conditions. The example is on the HI-STORM 100 cask at a proposed site containing four thousand such casks. The paper evaluates the risk to the public by determining the likelihood a welded stainless steel container will leak. In addition, the study addresses the risk at a site where 4,000 casks may be stored until the U.S. Department of Energy accepts the casks for placement in a repository. The methods used employ the PRODIGAL computer code to assess the probability of a faulty weld on a stainless steel-welded canister. These analyses are only the initial stages of a comprehensive risk study that the NRC is performing in support of its regulatory initiatives. (author)

  6. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    International Nuclear Information System (INIS)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-01-01

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10 3 and 6 x 10 4 rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10 4 rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10 5 rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance

  7. Project management for the Virginia power spent fuel storage project

    International Nuclear Information System (INIS)

    Smith, M.

    1992-01-01

    Like Duke Power, Virginia Power has been involved in spent fuel storage expansion studies for a long time - possibly a little longer than Duke Power. Virginia Power's initial studies date back to the late 70s and into the early 80s. Large variety of storage techniques are reviewed including reracking and transshipment. Virginia Power also considered construction a new spent fuel pool. This was one of the options that was considered early on since Virginia Power started this process before any dry storage techniques had been proven. Consolidation of spent fuel is something that was also studied. Finally, construction of dry storage facility was determined to be the technology of choice. They looked a large variety of dry storage technologies and eventually selected dry storage in metal casks at Surry. There are many of reasons why a utility may choose one technology over another. In Virginia Power's situation, additional storage was needed at Surry much earlier than at other utilities. Virginia Power was confronted with selecting a storage technique and having to be a leader in that it was the first U.S. utility to implement a dry storage system

  8. Device for sealing and shielding a nuclear fuel storage tank

    International Nuclear Information System (INIS)

    Masaki, Gengo.

    1975-01-01

    Object: To provide a shield device for opening and closing a great opening in a relay-storage-tank within a hot cell for temporarily storing a nuclear fuel, in which the device is simplified in construction and which can perform the opening and closing operation in simple, positive and quick manner. Structure: A biological shield is positioned upwardly of an opening of a nuclear fuel storage tank to render an actuator inoperative. A sealing plate, which is pivotally supported by a plurality of support rod devices from the biological shield for parallel movement with respect to the biological shield, comes in contact with a resilient seal disposed along the entire peripheral edge of the opening to form an air-tight seal therebetween. In order to release the opening, the actuator is first actuated and the end of the sealing plate is horizontally pressed by a piston rod thereof. Then, the sealing plate is moved along the line depicted by the end of the support rod in the support rod devices and as a consequence, the plate is moved away from the resilient seal in the peripheral edge of the opening. When a driving device is actuated to travel the plate along the aforesaid line while maintaining the condition as described, the biological device moves along the guide. (Kamimura, M.)

  9. Three-dimensional seismic analysis for spent fuel storage rack

    International Nuclear Information System (INIS)

    Lee, Gyu Mahn; Kim, Kang Soo; Park, Keun Bae; Park, Jong Kyun

    1998-01-01

    Time history analysis is usually performed to characterize the nonlinear seismic behavior of a spent fuel storage rack (SFSR). In the past, the seismic analyses of the SFSR were performed with two-dimensional planar models, which could not account for torsional response and simultaneous multi-directional seismic input. In this study, three-dimensional seismic analysis methodology is developed for the single SFSR using the ANSY code. The 3-D model can be used to determine the nonlinear behavior of the rack, i.e., sliding, uplifting, and impact evaluation between the fuel assembly and rack, and rack and the pool wall. This paper also reviews the 3-D modeling of the SFSR and the adequacy of the ANSYS for the seismic analysis. As a result of the adequacy study, the method of ANSYS transient analysis with acceleration time history is suitable for the seismic analysis of highly nonlinear structure such as an SFSR but it isn't appropriate to use displacement time history of seismic input. (author)

  10. Bases for extrapolating materials durability in fuel storage pools

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1994-12-01

    A major body of evidence indicates that zirconium alloys have the most consistent and reliable durability in wet storage, justifying projections of safe wet storage greater than 50 y. Aluminum alloys have the widest range of durabilities in wet storage; systematic control and monitoring of water chemistry have resulted in low corrosion rates for more than two decades on some fuels and components. However, cladding failures have occurred in a few months when important parameters were not controlled. Stainless steel is extremely durable when stress, metallurgical and water chemistry factors are controlled. LWR SS cladding has survived for 25 y in wet storage. However, sensitized, stressed SS fuels and components have seriously degraded in fuel storage pools (FSPs) at ∼ 30 C. Satisfactory durability of fuel assembly and FSP component materials in extended wet storage requires investments in water quality management and surveillance, including chemical and biological factors. The key aspect of the study is to provide storage facility operators and other decision makers a basis to judge the durability of a given fuel type in wet storage as a prelude to basing other fuel management plans (e.g. dry storage) if wet storage will not be satisfactory through the expected period of interim storage

  11. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    Energy Technology Data Exchange (ETDEWEB)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. [INEEL (US); Brey, R.F. [ISU (US); Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  12. Criticality and Its Uncertainty Analysis of Spent Fuel Storage Rack for Research Reactor

    International Nuclear Information System (INIS)

    Han, Tae Young; Park, Chang Je; Lee, Byung Chul

    2011-01-01

    For evaluating the criticality safety of spent fuel storage rack in an open pool type research reactor, a permissible upper limit of criticality should be determined. It can be estimated from the criticality upper limit presented by the regulatory guide and an uncertainty of criticality calculation. In this paper, criticalities for spent fuel storage rack are carried out at various conditions. The calculation uncertainty of MCNP system is evaluated from the calculation results for the benchmark experiments. Then, the upper limit of criticality is determined from the uncertainties and the calculated criticality of the spent fuel storage rack is evaluated

  13. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  14. 75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

    Science.gov (United States)

    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...

  15. Fuel storage tanks at FAA facilities : Order 1050.15A : executive summary.

    Science.gov (United States)

    1997-04-30

    The Federal Aviation Administration (FAA) has over 4,000 fuel storage tanks (FST) in its inventory. Most of these FSTs are underground storage tanks (UST) that contain fuel for emergency backup generators providing secondary power to air navigational...

  16. Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

    2014-05-15

    700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

  17. Bulk Fuel Storage and Delivery Systems Infrastructure Military Construction Requirements for Japan

    National Research Council Canada - National Science Library

    Padgett, Gary

    2000-01-01

    This report is one in a series that addresses the accuracy and reliability of maintenance, repair, and environmental and construction requirements for bulk fuel storage and delivery systems infrastructure...

  18. Spent fuel storage facility at science and technical center 'Sosny': Experience of ten years activity

    International Nuclear Information System (INIS)

    Chigrinov, S.; Goulo, V.; Lunev, A.; Belousov, N.; Salnikov, L.; Boiko, L.

    2000-01-01

    Spent fuel storage of the Academic Science and Technical Center in Minsk is in operation already more then 10 years. In the paper aspects of its design, operation practice, problems and decisions for future are discussed. (author)

  19. Cask operation and maintenance for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [International Atomic Energy Agency, Vienna (Austria)

    2004-07-01

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage.

  20. Cask operation and maintenance for spent fuel storage

    International Nuclear Information System (INIS)

    Lee, J.S.

    2004-01-01

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage

  1. Spent nuclear fuel storage: Legal, technical and political considerations

    International Nuclear Information System (INIS)

    Blake, E.L. Jr.; Buren, M.A.

    1994-01-01

    In 1982, Congress enacted the Nuclear Waste Policy Act (NWPA), assigning responsibility to the Department of Energy (DOE) for the development and implementation of a comprehensive national nuclear waste management program. The NWPA makes clear that the generators and owners of commercially-generated spent nuclear fuel (SNF) have the primary responsibility to provide for, and pay the costs of, the interim storage of such SNF until it is accepted by the DOE under the provisions of the NWPA. The shift in responsibility was expected to begin in 1998, the date specified in the NWPA and the DOE's contracts with the utilities, at which time the NWPA anticipated commencement of operations of a geologic repository and/or a monitored retrievable storage facility (MRS). Unfortunately, despite a mid-course correction to the NWPA mandated by Congress in 1987 in an effort to streamline and accelerate the program, DOE is way behind schedule. DOE's last published program schedule indicates the commencement of repository operations in 2010, a date many feel is overly optimistic. In repeated statements during the early 1990s, DOE sought to reassure utility companies and their regulatory commissions that it could still commence SNF acceptance in 1998 for storage at an MRS if such a facility were sited through a voluntary process by the end of 1992. That date has now come and gone. Although DOE is still nominally seeking a voluntary MRS host jurisdiction, the prospects for MRS operation by 1998 are dim. Putting aside for the moment the question of DOE's ability to bring the repository on line, the immediate problem facing domestic utilities is the need to augment their onsite SNF storage capacity. In addition to providing a brief overview of the Federal independent spent fuel storage installation (ISFSI) licensing process, the author provides some insight of what the real issues are in ISFSI licensing

  2. Seismic analysis of spent nuclear fuel storage racks

    International Nuclear Information System (INIS)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B.; Harstead, G.A.; Marquet, F.

    1996-01-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. No structural benefit from the BSS is assumed. This paper describes the methods used to perform seismic analysis of high density spent fuel storage racks. The sensitivity of important parameters such as the effect of variation of coefficients of friction between the rack legs and the pool floor and fuel loading conditions (consolidated and unconsolidated) are also discussed in the paper. Results of this study are presented. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, and the type of the seismic event. This paper presents several of the mathematical models usually used. Friction cannot be precisely predicted, so a range of friction coefficients is assumed. The range assumed for the analysis is 0.2 to 0.8. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are normally analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies

  3. Super-structure and building performance

    CSIR Research Space (South Africa)

    Van Wyk, Llewellyn V

    2010-11-01

    Full Text Available The super-structure consists predominantly of the load- and no-load-bearing walls-including all doors and windows and suspended floor slabs. The building envelope plays a significant role in the performance of a building, especially with regard...

  4. Manufacturing and Construction of Fresh Fuel Storage Rack for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this study is to provide the manufacturing and construction information regarding the Fresh Fuel Storage Rack (FFSR). The main function of a FFSR is to store and protect the 2 core new fuel assemblies for the operation of the research reactor. The fresh fuel assemblies are stored in a rack made of stainless steel and the storage rack is installed in the fresh fuel storage room. The fresh fuel facility provides fresh fuel assemblies with dry storage space. General design requirements of the fresh fuel storage facilities are given in the ANSI-57.3. Design, manufacturing, and construction of the fresh fuel storage rack are introduced. The analysis is performed to confirm the structural intensity of the fresh fuel storage rack under the seismic loads. The fresh fuel storage rack designed for storage of fresh fuel assemblies should be manufactured and installed with consideration of predicted number of fresh fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability.

  5. Current status of the first interim spent fuel storage facility in Japan

    International Nuclear Information System (INIS)

    Shinbo, Hitoshi; Kondo, Mitsuru

    2008-01-01

    In Japan, storage of spent fuels outside nuclear power plants was enabled as a result of partial amendments to the Nuclear Reactor Regulation Law in June 2000. Five months later, Mutsu City in Aomori Prefecture asked the Tokyo Electric Power Company (TEPCO) to conduct technical surveys on siting of the interim spent fuel storage facility (we call it 'Recyclable-Fuel Storage Center'). In April 2003, TEPCO submitted the report on siting feasibility examination, concluded that no improper engineering data for siting, construction of the facility will be possible from engineering viewpoint. Siting Activities for publicity and public acceptance have been continued since then. After these activities, Aomori Prefecture and Mutsu City approved siting of the Recyclable Fuel Storage Center in October 2005. Aomori Prefecture, Mutsu City, TEPCO and Japan Atomic Power Company (JAPC) signed an agreement on the interim spent fuel storage Facility. A month later, TEPCO and JAPC established Recyclable-Fuel Storage Company (RFS) in Mutsu City through joint capital investment, specialized in the first interim spent fuel storage Facility in Japan. In May 2007, we made an application for establishment permit, following safety review by regulatory authorities. In March 2008, we started the preparatory construction. RFS will safely store of spent fuels of TEPCO and JAPC until they will be reprocessed. Final storage capacity will be 5,000 ton-U. First we will construct the storage building of 3,000 ton-U to be followed by second building. We aim to start operation by 2010. (author)

  6. Manufacturing and Construction of Fresh Fuel Storage Rack for a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The objective of this study is to provide the manufacturing and construction information regarding the Fresh Fuel Storage Rack (FFSR). The main function of a FFSR is to store and protect the 2 core new fuel assemblies for the operation of the research reactor. The fresh fuel assemblies are stored in a rack made of stainless steel and the storage rack is installed in the fresh fuel storage room. The fresh fuel facility provides fresh fuel assemblies with dry storage space. General design requirements of the fresh fuel storage facilities are given in the ANSI-57.3. Design, manufacturing, and construction of the fresh fuel storage rack are introduced. The analysis is performed to confirm the structural intensity of the fresh fuel storage rack under the seismic loads. The fresh fuel storage rack designed for storage of fresh fuel assemblies should be manufactured and installed with consideration of predicted number of fresh fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability

  7. Microbial degradation processes in radioactive waste repository and in nuclear fuel storage areas

    International Nuclear Information System (INIS)

    Wolfram, J.H.; Rogers, R.D.; Gazso, L.G.

    1997-01-01

    The intent of the workshop organizers was to convene experts in the fields of corrosion and spent nuclear fuels. The major points which evolved from the interaction of microbiologists, material scientists, and fuel storage experts are as follows: Corrosion of basin components as well as fuel containers or cladding is occurring; Water chemistry monitoring, if done in the storage facility does not take into account the microbial component; Microbial influenced corrosion is an area that many have not considered to be an important contributor in the aging of metallurgical materials especially those exposed to a radiation field; Many observations indicate that there is a microbial or biological presence in the storage facilities but these observations have not been correlated with any deterioration or aging phenomena taking place in the storage facility; The sessions on the fundamentals of microbial influenced corrosion and biofilm pointed out that these phenomena are real, occurring on similar materials in other industries and probably are occurring in the wet storage of spent fuel; All agreed that more monitoring, testing, and education in the field of biological mediate processes be performed and financially supported; Loosing the integrity of fuel assemblies can only cause problems, relating to the future disposition of the fuel, safety concerns, and environmental issues; In other rad waste scenarios, biological processes may be playing a role, for instance in the mobility of radionuclides in soil, decomposition of organic materials of the rad waste, gas production, etc. The fundamental scientific presentations discussed the full gamut of microbial processes that relate to biological mediated effects on metallic and non-metallic materials used in the storage and containment of radioactive materials

  8. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    International Nuclear Information System (INIS)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

  9. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

  10. Waste management implications of irradiated nuclear fuel storage

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1977-01-01

    Irradiated nuclear fuel is stored underwater in large recirculating basins prior to reprocessing in chemical separations plants. A small number of the fuel rods contain minor defects which allow fission products to seep into the basin water. The predominant radionuclides leached into the water are dependent upon the decay time after removal from the reactor. Freshly discharged fuel releases short half-lived radioiodine which presents exposure and airborne release problems on a short-term basis but does not impose significant long-term waste management problems. After a reasonable decay period, the major radionuclides present are 137 Cs, 144 Ce, 90 Sr, 3 H, and 239 / 240 Pu. Each of the radionuclides imposes specific waste management problems which require individual consideration in their control, packaging and shipment, and permanent disposal. These considerations are reviewed in this paper for general industry practices and specific illustrative examples are presented. Control of radionuclide concentrations and water purity are normally achieved by filtration and ion exchange treatment. Cartridge-type filters are the most common but improved filtration techniques that minimize personnel exposure, particularly during filter replacements, are now being adopted. Mixed bed ion exchange resins are most commonly utilized for soluble radionuclide removal, particularly for basins filled with demineralized water. Cesium-specific exchange media are employed at basins where demineralized water is not employed; these media operate for very long periods of time since they are not depleted by the normal dissolved non-radioactive water impurities. The resins are either buried when depleted or regenerated and the regeneration solutions concentrated for burial. Resin run lengths are usually determined by ionic or radiochemical depletion of the resin or in some cases by limiting radionuclide concentrations specified by shipping regulations or established ion column dose rates

  11. 305 Building K basin mockup facility functions and requirements

    International Nuclear Information System (INIS)

    Steele, R.M.

    1994-01-01

    This document develops functions and requirements for installation and operation of a cold mockup test facility within the 305 Building. The test facility will emulate a portion of a typical spent nuclear fuel storage basin (e.g., 105-KE Basin) to support evaluation of equipment and processes for safe storage and disposition of the spent nuclear fuel currently within the K Basins

  12. Role of transportation in the utilities' management of spent fuel storage

    International Nuclear Information System (INIS)

    Newman, D.F.

    1985-01-01

    Additional spent fuel storage can be provided by using a combination of wet and dry storage technologies, with the technology or technologies used in any specific instance being determined by the particular circumstances involved. The capability for spent fuel storage at a reactor site can be enhanced using any one or a combination of the following: expansion of existing pool storage capacity; more efficient use of available capacity; and addition of an independent spent fuel storage installation (ISFSI). Each of these methods, which are described more fully below, have characteristics that may make them more or less suitable for use, depending on the nuclear power plant where they will be deployed, the magnitude of the need for additional storage, the utility's overall spent fuel management strategy, and other factors. 15 refs., 2 figs., 2 tabs

  13. Technical framework to facilitate foreign spent fuel storage and geologic disposal in Russia

    International Nuclear Information System (INIS)

    Jardine, L.J.; Halsey, W.G.; Cmith, C.F.

    2000-01-01

    The option of storage and eventual geologic disposal in Russia of spent fuel of US origin used in Taiwan provides a unique opportunity that can benefit many parties. Taiwan has a near term need for a spent fuel storage and geologic disposal solution, available financial resources, but limited prospect for a timely domestic solution. Russia has significant spent fuel storage and transportation management experience, candidate storage and repository sites, but limited financial resources available for their development. The US has interest in Taiwan energy security, national security and nonproliferation interests in Russian spent fuel storage and disposal and interest in the US origin fuel. While it is understood that such a project includes complex policy and international political issues as well as technical issues, the goal of this paper is to begin the discussion by presenting a technical path forward to establish the feasibility of this concept for Russia

  14. Dry spent fuel storage experience at overseas nuclear stations focus USA

    International Nuclear Information System (INIS)

    Bradley, T. L.; Kumar, S.; Marcelli, D. G.

    1997-01-01

    This paper provides a summary of US dry spent fuel storage experience, including application of this experience outside the United States. Background information on the US nuclear and spent fuel storage industry is provided as a basis for discussing the various types of options and systems available. An overview of technology options is presented, including systems being used and/or considered by the US government and private sector, as well as a discussion of overall system design, licensing and operation. Factors involved in selecting the best available technology option for a specific site or group of sites are presented, along with a typical timeline for project implementation. Cross-geographical use of technologies under different regulatory and technological regimes is also discussed. The paper concludes that dry storage is safe and reliable based on a successful ten year period. The information presented may be considered for use in the development of dry spent fuel storage in Korea and other countries. (author)

  15. Numerical simulation of ambient flow and thermal distributions in a spent fuel storage cask array

    International Nuclear Information System (INIS)

    Michener, T.; Trent, D.S.; Guttmann, J.; Bajwa, C.

    2001-01-01

    At the request of the U.S. Nuclear Regulatory Commission (USNRC), the staff at the Pacific Northwest National Laboratory (PNNL) analyzed the thermal performance of the Utah Private Fuel Storage (PFS) using the TEMPEST computational fluid dynamics software. A three-dimensional section of the PFS with a total of 20 casks was modeled to estimate the ambient flow and temperature distributions surrounding the casks. The purpose of this analysis was to compute the cask inlet vent air temperature to be used for boundary conditions in a detailed analysis of an individual Holtec Hi-Storm 100 cask using the COBRA-SFS (Spent Fuel Storage) thermal hydraulic computer software. (author)

  16. Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

    International Nuclear Information System (INIS)

    Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

    2011-01-01

    The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

  17. Operation and maintenance of spent fuel storage and transportation casks/containers

    International Nuclear Information System (INIS)

    2007-01-01

    Member States have a growing need for casks for spent fuel storage and transportation. A variety of casks has been developed and is in use at an increasing number of sites. This has resulted in an accumulation of experience that will provide valuable information for other projects in spent fuel management. This publication provides a comprehensive review of information on the cask operation and maintenance associated with spent fuel storage. It draws upon generic knowledge from industrial experience and applications and is intended to serve as a basis for better planning and implementation in future projects

  18. The NRC activities concerning Boraflex use in spent-fuel storage racks

    International Nuclear Information System (INIS)

    Kopp, L.I.

    1996-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has issued several generic communications to the nuclear industry identifying two issues with respect to using Boraflex in spent-fuel storage racks. The first issue related to gamma-radiation-induced shrinkage of Boraflex and the potential to develop tears or gaps in the material. This phenomenon is typically accounted for in criticality analyses of spent-fuel storage racks. The second issue concerned long-term Boraflex performance throughout the intended service life of the racks as a result of both gamma irradiation and exposure to the wet pool environment

  19. Seismic testing of the base-isolated PWR spent-fuel storage rack

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Tanaka, Mamoru; Nakamura, Masaaki; Tsujikura, Yonezo.

    1990-01-01

    The present paper aims to verify the seismic safety of the base-isolated spent-fuel storage rack. A series of seismic tests has been conducted using a three-dimensional shaking table. A sliding-type base-isolation system was employed for the prototype rack considering environmental conditions in an actual plant. A non linear seismic response analysis was also performed, and it is verified that the prototype of a base-isolated spent-fuel storage rack has a sufficient seismic safety margin for design seismic conditions from the viewpoint of seismic response. (author)

  20. Onsite dry spent-fuel storage: Becoming more of a reality

    International Nuclear Information System (INIS)

    1994-01-01

    An overview is presented of dry spent-fuel storage facilities operated at nuclear power plant sites in the USA. The experience of the utilities Virginia Power, Carolina Power and Light Company, Duke Power, Public Service Company of Colorado a Baltimore Gas and Electric is outlined. The spent fuel storage procedure using the Sierra Nuclear container system is described. Plans for the construction of additional storage facilities are mentioned. Dry stores are also operated at nuclear power plants that have been shut down. (J.B.). 1 fig

  1. Superconducting superstructure for the TESLA collider

    Energy Technology Data Exchange (ETDEWEB)

    Sekutowicz, J.; Tang, C. [DESY, MHF-SL, Hamburg (Germany); Ferrario, M. [INFN, Laboratori Nazionali di Frascati, Rome (Italy)

    1998-04-01

    The Authors discuss the new layout of a cavity chain (superstructure) allowing, the Authors hope, significant cost reduction of the RF system of both linacs of the TESLA linear collider. The proposed scheme increases the fill factor and thus makes an effective gradient of an accelerator higher. The authors present mainly computations that have performed up to now and which encouraged to order the copper model of the scheme, still keeping in mind that experiments with a beam will be necessary to prove if the proposed solution can be used for the acceleration.

  2. 76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    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...

  3. 78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    2013-12-06

    ...-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...

  4. 75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

    Science.gov (United States)

    2010-09-23

    ... Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective Date AGENCY: Nuclear... amended the NRC's spent fuel storage regulations at 10 CFR 72.214 to revise the NAC-MPC System listing to... configuration of the NAC-MPC storage system by the incorporation of a single closure lid with a welded closure...

  5. 30 CFR 75.1903 - Underground diesel fuel storage facilities and areas; construction and safety precautions.

    Science.gov (United States)

    2010-07-01

    ... areas; construction and safety precautions. 75.1903 Section 75.1903 Mineral Resources MINE SAFETY AND...; construction and safety precautions. (a) Permanent underground diesel fuel storage facilities must be— (1... with at least 240 pounds of rock dust and provided with two portable multipurpose dry chemical type...

  6. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

    ... Storage Casks: MAGNASTOR[supreg] System, Revision 2 AGENCY: Nuclear Regulatory Commission. ACTION: Direct... storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR[supreg] System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2 to Certificate of...

  7. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    International Nuclear Information System (INIS)

    Tyacke, M.J.; McConnell, J.W. Jr.; Ayers, A.L. Jr.; O'Connor, S.C.; Jankovich, J.P.

    1996-01-01

    The Idaho National Engineering Laboratory prepared a technical report for the Office of Nuclear Material Safety and Safeguards of the US Nuclear Regulatory Commission, entitled Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, NUREG/CR-6407. This paper provides the results of that report. It also presents the graded approach for classification of components used in transportation packagings and dry spent fuel storage systems. This approach provides a method for identifying the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements. The general types of transportation packagings and dry storage systems are identified. The methodology used in this paper is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This paper also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems, with a classification category assigned to each component. Several examples concerning the safety importance of components are presented

  8. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... Storage Casks: Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System AGENCY: Nuclear Regulatory... storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS[supreg] Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 11 to Certificate of...

  9. 75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ... Fuel Storage Casks: NAC-MPC System, Revision 6 AGENCY: Nuclear Regulatory Commission. ACTION: Direct...-MPC storage system as noted in Appendix B of the Technical Specifications (TS): Incorporation of a... include the following changes to the configuration of the NAC-MPC storage system as noted in Appendix B of...

  10. 75 FR 27401 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

    Science.gov (United States)

    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...

  11. Nonlinear analysis and evaluation of a reinforced concrete spent fuel storage pool for accidental thermal loads

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.

    1991-01-01

    A feasibility study was conducted for addition of consolidated fuel racks to an existing reinforced concrete spent fuel storage pool of a Mark I BWR plant. Nonlinear analysis of a detailed three-dimensional model of the fuel pool, considering cracking in concrete under gravity and thermal load conditions, showed that the pool has reserve capacities to carry the additional loads. (author)

  12. 75 FR 77017 - Nextera Energy Seabrook, LLC Seabrook Station Independent Spent Fuel Storage Installation; Exemption

    Science.gov (United States)

    2010-12-10

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-443, 72-63; NRC-2010-0381] Nextera Energy Seabrook, LLC Seabrook Station Independent Spent Fuel Storage Installation; Exemption 1.0 Background NextEra Energy Seabrook, LLC (NextEra, the licensee) is the holder of Facility Operating License No. NPF-86...

  13. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel storage rack

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mohd Fazli Zakaria; Zaredah Hashim; Ahmad Nabil Ab Rahim; Mohammad Suhaimi Kassim

    2010-01-01

    PUSPATI TRIGA Reactor (RTP) is a pool type research reactor with 1MW thermal power. It has been safely operated since 28 June 1982. During 28 years of safe operation, there are several systems and components of the RTP that have been maintained, repaired, upgraded and replaced in order to maintain its function and safety conditions. RTP has been proposed to be upgraded so that optimum operation of RTP could be achieved as well as fulfill the future needs. Thus, competencies and technical capabilities were needed to design and develop the reactor system. In the meantime, there is system or component need to be maintained such as fuel elements. Since early operation, most of the fuel elements still can be used and none of the fuel elements was replaced or sent for reprocessing and final disposal. Towards the power upgrading, preparation of spent fuel storage is needed for temporary storing of the fuels discharged from the reactor core. The spent fuel storage rack will be located in the spent fuel pool to accommodate the spent fuels before it is send to reprocessing or final disposal. This paper proposes the conceptual design of the spent fuel storage rack. The output of this paper focused on the physical and engineering design of the spent fuel storage. (author)

  14. 77 FR 7211 - Pacific Gas and Electric Company, Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2012-02-10

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 72-26; NRC-2011-0110] Pacific Gas and Electric Company...) issued NRC Materials License No. SNM-2511 to the Pacific Gas and Electric Company (PG&E) for the Diablo.... 5. TS 3.1.2, ``Spent Fuel Storage Cask (SFSC) Heat Removal System,''--revise to allow the HI-STORM...

  15. 78 FR 56944 - Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-09-16

    ... process waste at the Humboldt Bay ISFSI will not significantly affect the quality of the human environment... NUCLEAR REGULATORY COMMISSION [Docket No. 72-27; NRC-2011-0115] Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory Commission. ACTION...

  16. Applying x-ray digital imaging to the verification of cadmium in fuel-storage components

    International Nuclear Information System (INIS)

    Dabbs, R.D.; Cook, D.H.

    1997-01-01

    The High Flux Isotope Reactor utilizes large underwater fuel-storage arrays to stage irradiated fuel before it is shipped from the facility. Cadmium is required as a thermal neutron absorber in these fuel-storage arrays to produce an acceptable margin of nuclear subcriticality during both normal and off-normal operating conditions. Due to incomplete documentation from the time of their fabrication, the presence of cadmium within two stainless-steel parts of fuel-storage arrays must be experimentally verified before they are reused in new fuel-storage arrays. A cadmium-verification program has been developed in association with the Waste Examination and Assay Facility located at the Oak Ridge national Laboratory to nondestructively examine these older shroud assemblies. The program includes the following elements (1) x-ray analog imaging, (2) x-ray digital imaging, (3) prompt-gamma-ray spectroscopy measurements, and (4) neutron-transmission measurements. X-ray digital imaging utilizes an analog-to-digital convertor to record attenuated x-ray intensities observed on a fluorescent detector by a video camera. These x-ray intensities are utilized in expressions for cadmium thickness based upon x-ray attenuation theory

  17. IAEA Technical committee meeting on methods used in design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    Vitikainen, E.; Silfverberg, P.

    1985-01-01

    The meeting was held in Espoo, Finland and hosted by the Technical Research Centre of Finland (VTT), and was arranged to report and discuss design methods, licensing practise, operational experience as well as economic aspects connectied with spent fuel storage. This report contains session summaries by the session chairmen and the papers presented at the meeting

  18. Spent fuel storage practices and perspectives for WWER fuel in Eastern Europe

    International Nuclear Information System (INIS)

    Takats, F.

    1999-01-01

    In this lecture the general issues and options in spent fuel management and storage are reviewed. Quantities of spent fuel world-wide and spent fuel amounts in storage as well as spent fuel capacities are presented. Selected examples of typical spent fuel storage facilities are discussed. The storage technologies applied for WWER fuel is presented. Description of other relevant storage technologies is included

  19. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  20. 30 CFR 75.1912 - Fire suppression systems for permanent underground diesel fuel storage facilities.

    Science.gov (United States)

    2010-07-01

    ... Diesel-Powered Equipment § 75.1912 Fire suppression systems for permanent underground diesel fuel storage... system by a nationally recognized independent testing laboratory and appropriate for installation at a... recommended inspection and maintenance program and as required by the nationally recognized independent...

  1. Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

    2011-01-01

    for water chemistry control in research reactors; defines parameters recommended, techniques applicable, sampling procedures and sampling frequency to monitor water quality in RRs, and describes the importance of a quality assurance programme, and the implementation of a corrosion surveillance programme (CSP) as part of the water management programme. Whenever applicable, considerations are made for primary cooling system, spent fuel storage basins, secondary cooling system, emergency cooling systems, make-up systems and water reservoirs of RRs.

  2. Toolbox for super-structured and super-structure free multi-disciplinary building spatial design optimisation

    NARCIS (Netherlands)

    Boonstra, S.; van der Blom, K.; Hofmeyer, H.; Emmerich, M.T.M.; van Schijndel, A.W.M.; de Wilde, P.

    2018-01-01

    Multi-disciplinary optimisation of building spatial designs is characterised by large solution spaces. Here two approaches are introduced, one being super-structured and the other super-structure free. Both are different in nature and perform differently for large solution spaces and each requires

  3. Facility effluent monitoring plan for K area spent fuel storage basin

    International Nuclear Information System (INIS)

    Hunacek, G.S.

    1996-01-01

    A facility effluent monitoring plan is required by the U.S. Department of Energy in DOE Order 5400. 1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document was prepared using the specific guidelines identified in WHC-EP-0438-1, A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, and assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the second revision to the original annual report. Long-range integrity of the effluent monitoring system shall be ensured with updates of this report whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years

  4. Enhanced thermophysical properties via PAO superstructure

    Science.gov (United States)

    Pournorouz, Zahra; Mostafavi, Amirhossein; Pinto, Aditya; Bokka, Apparao; Jeon, Junha; Shin, Donghyun

    2017-01-01

    For the last few years, molten salt nanomaterials have attracted many scientists for their enhanced specific heat by doping a minute concentration of nanoparticles (up to 1% by weight). Likewise, enhancing the specific heat of liquid media is important in many aspects of engineering such as engine oil, coolant, and lubricant. However, such enhancement in specific heat was only observed for molten salts, yet other engineering fluids such as water, ethylene glycol, and oil have shown a decrease of specific heat with doped nanoparticles. Recent studies have shown that the observed specific heat enhancement resulted from unique nanostructures that were formed by molten salt molecules when interacting with nanoparticles. Thus, such enhancement in specific heat is only possible for molten salts because other fluids may not naturally form such nanostructures. In this study, we hypothesized such nanostructures can be mimicked through in situ formation of fabricated nano-additives, which are putative nanoparticles coated with useful organic materials (e.g., polar-group-ended organic molecules) leading to superstructures, and thus can be directly used for other engineering fluids. We first applied this approach to polyalphaolefin (PAO). A differential scanning calorimeter (DSC), a rheometer, and a customized setup were employed to characterize the heat capacity, viscosity, and thermal conductivity of PAO and PAO with fabricated nano-additives. Results showed 44.5% enhanced heat capacity and 19.8 and 22.98% enhancement for thermal conductivity and viscosity, respectively, by an addition of only 2% of fabricated nanostructures in comparison with pure PAO. Moreover, a partial melting of the polar-group-ended organic molecules was observed in the first thermal cycle and the peak disappeared in the following cycles. This indicates that the in situ formation of fabricated nano-additives spontaneously occurs in the thermal cycle to form nanostructures. Figure of merit analyses have

  5. Enhanced thermophysical properties via PAO superstructure.

    Science.gov (United States)

    Pournorouz, Zahra; Mostafavi, Amirhossein; Pinto, Aditya; Bokka, Apparao; Jeon, Junha; Shin, Donghyun

    2017-12-01

    For the last few years, molten salt nanomaterials have attracted many scientists for their enhanced specific heat by doping a minute concentration of nanoparticles (up to 1% by weight). Likewise, enhancing the specific heat of liquid media is important in many aspects of engineering such as engine oil, coolant, and lubricant. However, such enhancement in specific heat was only observed for molten salts, yet other engineering fluids such as water, ethylene glycol, and oil have shown a decrease of specific heat with doped nanoparticles. Recent studies have shown that the observed specific heat enhancement resulted from unique nanostructures that were formed by molten salt molecules when interacting with nanoparticles. Thus, such enhancement in specific heat is only possible for molten salts because other fluids may not naturally form such nanostructures. In this study, we hypothesized such nanostructures can be mimicked through in situ formation of fabricated nano-additives, which are putative nanoparticles coated with useful organic materials (e.g., polar-group-ended organic molecules) leading to superstructures, and thus can be directly used for other engineering fluids. We first applied this approach to polyalphaolefin (PAO). A differential scanning calorimeter (DSC), a rheometer, and a customized setup were employed to characterize the heat capacity, viscosity, and thermal conductivity of PAO and PAO with fabricated nano-additives. Results showed 44.5% enhanced heat capacity and 19.8 and 22.98% enhancement for thermal conductivity and viscosity, respectively, by an addition of only 2% of fabricated nanostructures in comparison with pure PAO. Moreover, a partial melting of the polar-group-ended organic molecules was observed in the first thermal cycle and the peak disappeared in the following cycles. This indicates that the in situ formation of fabricated nano-additives spontaneously occurs in the thermal cycle to form nanostructures. Figure of merit analyses have

  6. Information on the feasibility study for the reracking in the fuel storage pools of the Juragua Nuclear Power Plant

    International Nuclear Information System (INIS)

    Rodriguez, J.M.; Rodriguez, I.; Lopez, D.; Guerra, R.; Rodriguez, M.; Garcia, F.

    1995-01-01

    During 1993, in the Juragua Nuclear Power Plants as engineering evaluation programme was initiated in the storage area of irradiated nuclear fuel, where work in order to determine the feasibility of capacity increase for storage of irradiated nuclear fuel at the fuel storage pools using poisoned compact close racks instead of the originally designed racks. The feasibility study is a fundamental activity of this programme for the 1994-1995 period. According to this study the prospects of assimilation of compact storage conditions in the fuel storage pools in unit number one and prolonged fuel storage pool are investigated

  7. ELABORATION OF THE SUPERSTRUCTURE OF THE BULGARIAN HOTEL INDUSTRY

    Directory of Open Access Journals (Sweden)

    Tanya Dabeva

    2010-12-01

    Full Text Available The purpose of the paper is to suggest some terms of reference for the elaboration of the superstructure of the Bulgarian hotel industry. The suggestions are based on the analysis of the superstructure at present. Its capacity, main types of accommodation establishments and their distribution by categories and territory are examined. An analysis is made of some resulting indices such as overnight stays and average stay. The main positive, respectively negative features of the superstructure have been analyzed. Concrete terms of reference have been given in order to overcome the problems specified.

  8. Experience with the licensing of the interim spent fuel storage facility modification

    International Nuclear Information System (INIS)

    Bezak, S.; Beres, J.

    1999-01-01

    After political and economical changes in the end of eighties, the utility operating the nuclear power plants in the Slovak Republic (SE, a.s.) decided to change the original scheme of the back-end of the nuclear fuel cycle; instead of reprocessing in the USSR/Russian Federation spent fuel will be stored in an interim spent fuel storage facility until the time of the final decision. As the best solution, a modification of the existing interim spent fuel storage facility has been proposed. Due to lack of legal documents for this area, the Regulatory Authority of the Slovak Republic (UJD SR) performed licensing procedures of the modification on the basis of recommendations by the IAEA, the US NRC and the relevant parts of the US CFR Title 10. (author)

  9. Nuclear material control and accountancy in a spent fuel storage ponds

    International Nuclear Information System (INIS)

    Gurle, P.; Zhabo, Dgh.

    1999-01-01

    The spent fuel storage ponds of a large reprocessing plant La Hague in France are under safeguards by means of a wide range of techniques currently used. These techniques include the nuclear material accountancy an containment/surveillance (C/S). Nondestructive assay, design information verification, and authentication of equipment provided by the operator are also implemented. Specific C/S equipment including video surveillance and unattended radiation monitoring have been developed and implemented in a spent fuel pond of La Hague. These C/S systems named EMOSS and CONSULHA with high degree of reliability and conclusiveness provide the opportunity to improve the efficiency of safeguards, particularly as related to spent fuel storage areas where the accountancy is verified by item counting [ru

  10. A systems evaluation model for selecting spent nuclear fuel storage concepts

    International Nuclear Information System (INIS)

    Postula, F.D.; Finch, W.C.; Morissette, R.P.

    1982-01-01

    This paper describes a system evaluation approach used to identify and evaluate monitored, retrievable fuel storage concepts that fulfill ten key criteria for meeting the functional requirements and system objectives of the National Nuclear Waste Management Program. The selection criteria include health and safety, schedules, costs, socio-economic factors and environmental factors. The methodology used to establish the selection criteria, develop a weight of importance for each criterion and assess the relative merit of each storage system is discussed. The impact of cost relative to technical criteria is examined along with experience in obtaining relative merit data and its application in the model. Topics considered include spent fuel storage requirements, functional requirements, preliminary screening, and Monitored Retrievable Storage (MRS) system evaluation. It is concluded that the proposed system evaluation model is universally applicable when many concepts in various stages of design and cost development need to be evaluated

  11. Engineering program in order to increase the irradiated fuel storage capacity in pool facilities of Juragua

    International Nuclear Information System (INIS)

    Rodriguez R, J.

    1996-01-01

    In 1993, a technical program in the spent fuel storage area of Nuclear Plant Juragua was launched. Such a program tries to carry out an engineering assessment of the possibility of increasing the spent fuel storage capacity in pool storage facilities by using high density racks (re-racking) instead of the original (non-compact) ones. The purpose of the above-mentioned program is to evaluate possible solutions that can be applied to the construction works prior to plant operation. The first stage of the program for the 1994-95 period is an ongoing Engineering-Economic Feasibility Study (EEFS), which endeavors to examine the capabilities of the reloading pool in Unit-1 Reactor building and long-term storage pool in auxiliary building in high density storage conditions. Technical details of the EEFS and reached results and difficulties are described. (author). 5 refs., 2 figs

  12. Design ampersand operational experience of the NUHOMS reg-sign-24P spent fuel storage system

    International Nuclear Information System (INIS)

    McConaghy, W.J.; Lehnert, R.A.; Rasmussen, R.W.

    1991-01-01

    The NUHOMS reg-sign Spent Fuel Storage System provides a safe and economical method for the dry storage of spent fuel assemblies either at an at-reactor Independent Spent Fuel Storage Installation (ISFSI) or at a centralized away-from-reactor (AFR) storage facility. The system consists of three major safety related components: a dry shielded canister (DSC) which provides a high integrity containment boundary and a controlled storage environment for the fuel; a reinforced concrete horizontal storage module (HSM) which houses the stored DSC and provides radiation shielding, protection against natural phenomena, and an efficient means for decay heat removal; and a transfer cask which provides for the safe shielded transfer of the DSC from the plant spent fuel pool to the HSM. The NUHOMS reg-sign system is designed and licensed to the requirements of 10 CFR 72 and ANS/ANSI 57.9 for ISFSIs

  13. Spent fuel storage capacities. An update of DOE/RL-84-1

    International Nuclear Information System (INIS)

    1985-10-01

    Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The resport also assesses the possible impacts of increased fuel exposure and spent fuel transhipment on the requirements for additional storage capacity

  14. Spent fuel storage requirements: the need for away-from-reactor storage

    International Nuclear Information System (INIS)

    1980-01-01

    The analyses of on-site storage capabilities of domestic utilities and estimates of timing and magnitude of away-from-reactor (AFR) storage requirements were presented in the report DOE/ET-0075 entitled Spent Fuel Storage Requirements: The Need For Away-From-Reactor Storage published in February 1979 by the US Department of Energy. Since utility plans and requirements continue to change with time, a need exists to update the AFR requirements estimates as appropriate. This short report updates the results presented in DOE/ET-0075 to reflect recent data on reactor operations and spent fuel storage. In addition to the updates of cases representing the range of AFR requirements in DOE/ET-0075, new cases of interest reflecting utility and regulatory trends are presented

  15. Integrated management platform of nuclear fuel storage and transportation based on RFID

    International Nuclear Information System (INIS)

    Song Yafeng; Ma Yanqin; Chen Liyu; Jiang Yong; Wu Jianlei; Yang Haibo; Zhang Haiyan

    2012-01-01

    This paper describes integrated system model to improve work efficiency and optimize control measures of nuclear fuel storage and transportation, RFID and information integration technology is introduced, traditional management processes are innovated in data acquisition and monitoring fields as well, system solutions and design model are given by emphasizing on the following key technologies: cascade protection of information system, security protocol of RFID information, algorithm of collision. (authors)

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

    International Nuclear Information System (INIS)

    1997-07-01

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

  17. Special equipment support the fuel storage; Equipos especiales para apoyos al almacenamiento de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Vega, M. E.

    2014-10-01

    In the current juncture one of the keys to any company that works in a market that is as demanding as the nuclear, is its ability to developed new technological products that they can adapt to the different special situations/needs of nuclear Power Plants during their operating life. As an example, below are some of the specialized equipment that ENSA has been developing for more than thirty years that has been doing work in the area of fuel storage. (Author)

  18. Seismic evaluation of BWR spent fuel storage racks using actual damping by vibration test in water

    International Nuclear Information System (INIS)

    Yamasaki, Hiroto; Iwakura, Shigeyoshi; Imaoka, Tetsuo; Okumura, Kazue; Orita, Syuichi; Namita, Yoshio

    2010-01-01

    Damping value for BWR spent fuel storage racks has been used 1 percent damping, which is applied to welded steel structures in air as defined JEAG4601. However, it is considered that the actual damping is higher than that of the above mentioned, because of its underwater installation. This report shows the actual damping value of the Check Arrayed Rack by vibration test in water and Evaluation by the analysis of rack using actual damping. (author)

  19. Fuel Assemblies Thermal Analysis in the New Spent Fuel Storage Facility at Inshass Site

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, Ahmed

    1999-01-01

    New Wet Storage Facility (NSF) is constructed at Inshass site to solve the problem of spent fuel storage capacity of ETRR-1 reactor . The Engineering Safety Heat Transfer Features t hat characterize the new facility are presented. Thermal analysis including different scenarios of pool heat load and safety limits are discussed . Cladding temperature limit during handling and storage process are specified for safe transfer of fuel

  20. Problem of spent fuel storage from commercial nuclear reactors in USA

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    1996-01-01

    The problem on spent fuel storage in the USA is considered. According to the law and agreement, concluded with electrical companies, the USA should begin to receive the spent fuel from commercial reactors in 1998, however they are not ready for it. The consortium for constructing a centralized storage, financed from private sources for its temporary disposition is established recently. The spent fuel receipt is planned for 2002

  1. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    International Nuclear Information System (INIS)

    McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J.

    1996-02-01

    This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned

  2. Spent fuel storage requirements. An update of DOE/RL-83-1

    International Nuclear Information System (INIS)

    1984-05-01

    Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information voluntarily supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) Federal Interim Storage (FIS) Program and the spent fuel research, development, and demonstration (RD and D) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. Since the planning needs of the CSFM program focus on the near-term management of spent fuel inventories from commercial nuclear power reactors, the estimates in this report cover the ten-year period from the present through 1983. The report also assesses the possible impacts of using various concepts to reduce the requirements for additional storage capacity

  3. Czech interim spent fuel storage facility: operation experience, inspections and future plans

    International Nuclear Information System (INIS)

    Fajman, V.; Bartak, L.; Coufal, J.; Brzobohaty, K.; Kuba, S.

    1999-01-01

    The paper describes the situation in the spent fuel management in the Czech Republic. The interim Spent Fuel Storage Facility (ISFSF) at Dukovany, which was commissioned in January 1997 and is using dual transport and storage CASTOR - 440/84 casks, is briefly described. The authors deal with their experience in operating and inspecting the ISFSF Dukovany. The structure of the basic safety document 'Limits and Conditions of Normal Operation' is also mentioned, including the experience of the performance. The inspection activities focused on permanent checking of the leak tightness of the CASTOR 440/84 casks, the maximum cask temperature and inspections monitoring both the neutron and gamma dose rate as well as the surface contamination. The results of the inspections are mentioned in the presentation as well. The operator's experience with re-opening partly loaded and already dried CASTOR-440/84 cask, after its transport from NPP Jaslovske Bohunice to the NPP Dukovany is also described. The paper introduces briefly the concept of future spent fuel storage both from the NPP Dukovany and the NPP Temelin, as prepared by the CEZ. The preparatory work for the Central Interim Spent Nuclear Fuel Storage Facility (CISFSF) in the Czech Republic and the information concerning the planned storage technology for this facility is discussed in the paper as well. The authors describe the site selection process and the preparatory steps concerning new spent fuel facility construction including the Environmental Impact Assessment studies. (author)

  4. Spent fuel storage requirements. An update of DOE/RL-85-2

    International Nuclear Information System (INIS)

    1986-10-01

    Utility projections of spent fuel storage capacities indicate that some commercial light water reactors (LWRs) have inadequate capacity to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The report also assesses the possible impacts of increased fuel exposure and spent fuel transshipment on the requirements for additional storage capacity

  5. 78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

    Science.gov (United States)

    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...

  6. Integral ceramic superstructure evaluation using time domain optical coherence tomography

    Science.gov (United States)

    Sinescu, Cosmin; Bradu, Adrian; Topala, Florin I.; Negrutiu, Meda Lavinia; Duma, Virgil-Florin; Podoleanu, Adrian G.

    2014-02-01

    Optical Coherence Tomography (OCT) is a non-invasive low coherence interferometry technique that includes several technologies (and the corresponding devices and components), such as illumination and detection, interferometry, scanning, adaptive optics, microscopy and endoscopy. From its large area of applications, we consider in this paper a critical aspect in dentistry - to be investigated with a Time Domain (TD) OCT system. The clinical situation of an edentulous mandible is considered; it can be solved by inserting 2 to 6 implants. On these implants a mesostructure will be manufactured and on it a superstructure is needed. This superstructure can be integral ceramic; in this case materials defects could be trapped inside the ceramic layers and those defects could lead to fractures of the entire superstructure. In this paper we demonstrate that a TD-OCT imaging system has the potential to properly evaluate the presence of the defects inside the ceramic layers and those defects can be fixed before inserting the prosthesis inside the oral cavity. Three integral ceramic superstructures were developed by using a CAD/CAM technology. After the milling, the ceramic layers were applied on the core. All the three samples were evaluated by a TD-OCT system working at 1300 nm. For two of the superstructures evaluated, no defects were found in the most stressed areas. The third superstructure presented four ceramic defects in the mentioned areas. Because of those defects the superstructure may fracture. The integral ceramic prosthesis was send back to the dental laboratory to fix the problems related to the material defects found. Thus, TD-OCT proved to be a valuable method for diagnosing the ceramic defects inside the integral ceramic superstructures in order to prevent fractures at this level.

  7. K Basin safety analysis

    International Nuclear Information System (INIS)

    Porten, D.R.; Crowe, R.D.

    1994-01-01

    The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall

  8. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  9. Selection of away-from-reactor facilities for spent fuel storage. A guidebook

    International Nuclear Information System (INIS)

    2007-09-01

    This publication aims to provide information on the approaches and criteria that would have to be considered for the selection of away-from-reactor (AFR) type spent fuel storage facilities, needs for which have been growing in an increasing number of Member States producing nuclear power. The AFR facilities can be defined as a storage system functionally independent of the reactor operation providing the role of storage until a further destination such as a disposal) becomes available. Initially developed to provide additional storage space for spent fuel, some AFR storage options are now providing additional spaces for extended storage of spent fuel with a prospect for long term storage, which is becoming a progressive reality in an increasing number of Member States due to the continuing debate on issues associated with the endpoints for spent fuel management and consequent delays in the implementation of final steps, such as disposal. The importance of AFR facilities for storage of spent fuel has been recognized for several decades and addressed in various IAEA publications in the area of spent fuel management. The Guidebook on Spent Fuel Storage (Technical Reports Series No. 240 published in 1984 and revised in 1991) discusses factors to be considered in the evaluation of spent fuel storage options. A technical committee meeting (TCM) on Selection of Dry Spent Fuel Storage Technologies held in Tokyo in 1995 also deliberated on this issue. However, there has not been any stand-alone publication focusing on the topic of selection of AFR storage facilities. The selection of AFR storage facilities is in fact a critical step for the successful implementation of spent fuel management programmes, due to the long operational periods required for storage and fuel handling involved with the additional implication of subsequent penalties in reversing decisions or changing the option mid-stream especially after the construction of the facility. In such a context, the long

  10. Department of Energy report on fee for spent nuclear fuel storage and disposal services

    International Nuclear Information System (INIS)

    1980-10-01

    Since the July 1978 publication of an estimated fee for storage and disposal, several changes have occurred in the parameters which impact the spent fuel fee. DOE has mounted a diversified program of geologic investigations that will include locating and characterizing a number of potential repository sites in a variety of different geologic environments with diverse rock types. As a result, the earliest operation date of a geologic repository is now forecast for 1997. Finally, expanded spent fuel storage capabilities at reactors have reduced the projected quantities of fuel to be stored and disposed of. The current estimates for storage and disposal are presented. This fee has been developed from DOE program information on spent fuel storage requirements, facility availability, facility cost estimates, and research and development programs. The discounted cash flow technique has used the most recent estimates of cost of borrowing by the Federal Government. This estimate has also been used in calculating the Federal charge for uranium enrichment services. A prepayment of a percentage of the storage portion of the fee is assumed to be required 5 years before spent fuel delivery. These funds and the anticipated $300 million in US Treasury borrowing authority should be sufficient to finance the acquisition of storage facilities. Similarly, a prepayment of a percentage of the disposal portion would be collected at the same time and would be used to offset disposal research and development expenditures. The balance of the storage and disposal fees will be collected upon spent fuel delivery. If disposal costs are different from what was estimated, there will be a final adjustment of the disposal portion of the fee when the spent fuel is shipped from the AFR for permanent disposal. Based on current spent fuel storage requirements, at least a 30 percent prepayment of the fee will be required

  11. A nuclear fuel cycle system dynamic model for spent fuel storage options

    International Nuclear Information System (INIS)

    Brinton, Samuel; Kazimi, Mujid

    2013-01-01

    Highlights: • Used nuclear fuel management requires a dynamic system analysis study due to its socio-technical complexity. • Economic comparison of local, regional, and national storage options is limited due to the public financial information. • Local and regional options of used nuclear fuel management are found to be the most economic means of storage. - Abstract: The options for used nuclear fuel storage location and affected parameters such as economic liabilities are currently a focus of several high level studies. A variety of nuclear fuel cycle system analysis models are available for such a task. The application of nuclear fuel cycle system dynamics models for waste management options is important to life-cycle impact assessment. The recommendations of the Blue Ribbon Committee on America’s Nuclear Future led to increased focus on long periods of spent fuel storage [1]. This motivated further investigation of the location dependency of used nuclear fuel in the parameters of economics, environmental impact, and proliferation risk. Through a review of available literature and interactions with each of the programs available, comparisons of post-reactor fuel storage and handling options will be evaluated based on the aforementioned parameters and a consensus of preferred system metrics and boundary conditions will be provided. Specifically, three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module (WMM) which provides an easy to use interface for education on fuel cycle waste management economic impacts. Initial results of baseline cases point to positive benefits of regional storage locations with local regional storage options continuing to offer the lowest cost

  12. Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Improvement of numerical simulation methods on safety assessment of the spent fuel storage facility is one of main objectives of JNES activities. For the thermal and structural analyses, the radiative heat transfer analysis code S-FOKS has been developed to reduce computing time and to avoid using large memory area. In order to simulate the specular reflection, a new model (called 'model-2') is planned to install to S-FOKS code. The theoretical values with the specular reflection in simple geometry were lead to verify S-FOKS model-2. (author)

  13. Utilization of the NFS West Valley Installation for spent fuel storage

    International Nuclear Information System (INIS)

    MacDonald, R.W.

    1978-04-01

    Several thousand MT of capacity of AFR storage will be required in the 1980's. The pool at NFS has capacity for an additional 60 MT of BWR fuel or 150 MT of PWR assemblies. Zircaloy-clad LWR fuel can be stored in pools for up to 100 years. Environmental effects are discussed. Expansion of the pool capacity for as much as 1000 MT more, either by using more compact storage racks or constructing a new pool or an independent pool, is considered. Some indication of the environmental impacts of expanded fuel storage capacity at West Valley is offered by experience at Barnwell

  14. A Study on Criticality Safety Parameters of New Fuel Storage Rack Design

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Joon Gi; Kim, Hyeong Heon [KEPCO E and C, Daejeon (Korea, Republic of)

    2012-05-15

    The dry new fuel storage rack shall maintain the subcritical condition (i.e., k-eff < 0.95) when fully flooded with water and the k-eff will not exceed 0.98 even assuming that the optimum moderation causes the highest reactivity. Thus, the design parameters of the new fuel rack are determined optimally by considering both the full density water flooding condition and the optimum moderation condition. The behavior of the keff as the variation of design parameter of the new fuel rack was investigated as the function of the moderating water density

  15. Uses of the waste heat from the interim fuel storage facility

    International Nuclear Information System (INIS)

    Wehrum, A.

    It was the objective of this study to investigate the possibilities of a convenient use of the waste heat from the designed interim fuel storage at Ahaus. In this sense the following possibilities have been investigated: district heating, heat for industrial processes, fish-production, green house-heating, production of methane from original waste, agrotherm (agricultur field heating). It has been shown, that an economical behaviour for nearly all variations is not given without the financial help of the government, because of the high costs for heat transport and out-put. The most economical project is the intensive fish production plant. (orig.) [de

  16. The Spent Fuel Management in Finland and Modifications of Spent Fuel Storages

    International Nuclear Information System (INIS)

    Maaranen, Paeivi

    2014-01-01

    The objective of this presentation is to share the Finnish regulator's (STUK) experiences on regulatory oversight of the enlargement of a spent fuel interim storage. An overview of the current situation of spent fuel management in Finland will also be given. In addition, the planned modifications and requirements set for spent fuel storages due to the Fukushima accident are discussed. In Finland, there are four operating reactors, one under construction and two reactors that have a Council of State's Decision-in-Principle to proceed with the planning and licensing of a new reactor. In Olkiluoto, the two operating ASEA-Atom BWR units and the Areva EPR under construction have a shared interim storage for the spent fuel. The storage was designed and constructed in 1980's. The option for enlarging the storage was foreseen in the original design. Considering three operating units to produce their spent fuel and the final disposal to begin in 2022, extra space in the spent fuel storage is estimated to be needed in around 2014. The operator decided to double the number of the spent fuel pools of the storage and the construction began in 2010. The capacity of the enlarged spent fuel storage is considered to be sufficient for the three Olkiluoto units. The enlargement of the interim storage was included in Olkiluoto NPP 1 and 2 operating license. The licensing of the enlargement was conducted as a major plant modification. The operator needed the approval from STUK to conduct the enlargement. Prior to the construction of this modification, the operator was required to submit the similar documentation as needed for applying for the construction license of a nuclear facility. When conducting changes in an old nuclear facility, the new safety requirements have to be followed. The major challenge in the designing the enlargement of the spent fuel storage was to modify it to withstand a large airplane crash. The operator chose to cover the pools with protecting slabs and also to

  17. Expansion of storage capacity of interim spent fuel storage (MSVP) Bohunice

    International Nuclear Information System (INIS)

    Pilat, P.; Fridrich, V.

    2005-01-01

    This article describes modifications of Interim spent fuel storage, performed with aim of storage capacity expansion, seismic stability enhancement, and overall increase of service life as well as assuring of MSVP safe operation. Uniqueness of adopted technical solutions is based upon the fact that mentioned innovations and modifications were performed without any changes, neither in ground plan nor architecture of MSVP structure. It also important to mention that all modifications were performed during continual operation of MSVP without any breaks of limits or operational regulations. Reconstruction and innovation of existing construction and technological systems of MSVP has assured required quality standard comparable with actual trends. (authors)

  18. Licensing of spent fuel storage facility including its physical protection in the Czech Republic

    International Nuclear Information System (INIS)

    Fajman, V.; Sedlacek, J.

    1992-01-01

    The current spent fuel management policies as practised in the Czech Republic are described, and the conception of the fuel cycle back end is outlined. The general principles and the legislative framework are explained of the licensing process concerning spent fuel interim storage facilities, including the environmental impact assessment component. The history is outlined of the licensing process for the spent fuel storage facility at the Dukovany NPP site, including the licensing of the transport and storage cask. The basic requirements placed on the physical safeguarding of the facility and on the licensing process are given. (J.B.). 13 refs

  19. An effective and practical fire-protection system. [for aircraft fuel storage and transport

    Science.gov (United States)

    Mansfield, J. A.; Riccitiello, S. R.; Fewell, L. L.

    1975-01-01

    A high-performance sandwich-type fire protection system comprising a steel outer sheath and insulation combined in various configurations is described. An inherent advantage of the sheath system over coatings is that it eliminates problems of weatherability, materials strength, adhesion, and chemical attack. An experimental comparison between the protection performance of state-of-the-art coatings and the sheath system is presented, with emphasis on the protection of certain types of steel tanks for fuel storage and transport. Sheath systems are thought to be more expensive than coatings in initial implementation, although they are less expensive per year for sufficiently long applications.

  20. Shielding analysis of the LMR in-vessel fuel storage experiments

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    1994-01-01

    The In-Vessel Fuel Storage (IVFS) experiments analyzed in this paper were conducted at the Oak Ridge National Laboratory's Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present paper describes the 2- and 3-D calculations and results corresponding to a limited subset of those IVFS experiments in which the US LMR program had a particular interest

  1. Shielding calculation techniques used in the design of fuel storage systems

    International Nuclear Information System (INIS)

    Wang, S.S.; Massey, J.V.

    1986-01-01

    This paper addresses the shielding design and analysis of a concrete modular spent fuel storage system. Particular attention is given to comparing various computation techniques in determining bulk shielding thickness, and also in dealing with the radiation streaming effect through the air exit penetration openings in the module. Three computer codes QADMOD, ANISN, and DOT-IV were used to solve the same problem. In addition, hand albedo calculation were done to augment the result of the QADMOD calculation to properly deal with the surface scattering

  2. A Study on Criticality Safety Parameters of New Fuel Storage Rack Design

    International Nuclear Information System (INIS)

    Ahn, Joon Gi; Kim, Hyeong Heon

    2012-01-01

    The dry new fuel storage rack shall maintain the subcritical condition (i.e., k-eff < 0.95) when fully flooded with water and the k-eff will not exceed 0.98 even assuming that the optimum moderation causes the highest reactivity. Thus, the design parameters of the new fuel rack are determined optimally by considering both the full density water flooding condition and the optimum moderation condition. The behavior of the keff as the variation of design parameter of the new fuel rack was investigated as the function of the moderating water density

  3. Technology transfer and design conversion of a dry spent fuel storage system in Ukraine

    International Nuclear Information System (INIS)

    Peacock, R.C.; Marcelli, D.G.

    1998-01-01

    A number of unique issues surfaced in the technology transfer and design conversion of a US dry spent fuel storage technology in Ukraine. Unique challenges were encountered in the areas of nuclear design conversion, technical codes and standards, material selection and qualification, fabrication, construction and testing, quality assurance, documentation, and translation and verification processes. Technology transfer and design conversion were undertaken for both concrete and steel components for the project. The overall effort presented significant technical and cultural challenges to both the US and Ukrainian side, but technical exchange and design improvements to achieve a common goal have been reached. (author)

  4. Evaluation of AMPX-KENO benchmark calculations for high-density spent fuel storage racks

    International Nuclear Information System (INIS)

    Turner, S.E.; Gurley, M.K.

    1981-01-01

    The AMPX-KENO computer code package is commonly used to evaluate criticality in high-density spent fuel storage rack designs. Consequently, it is important to know the reliability that can be placed on such calculations and whether or not the results are conservative. This paper evaluates a series of AMPX-KENO calculations which have been made on selected critical experiments. The results are compared with similar analyses reported in the literature by the Oak Ridge National Laboratory and BandW. 8 refs

  5. Verification of criticality safety in on-site spent fuel storage systems

    International Nuclear Information System (INIS)

    Rasmussen, R.W.

    1989-01-01

    On February 15, 1984, Duke Power Company received approval for a two-region, burnup credit, spent fuel storage rack design at both Units 1 and 2 of the McGuire Nuclear Station. Duke also hopes to obtain approval by January of 1990 for a dry spent fuel storage system at the Oconee Nuclear Station, which will incorporate the use of burnup credit in the criticality analysis governing the design of the individual storage units. While experiences in burnup verification for criticality safety for their dry storage system at Oconee are in the future, the methods proposed for burnup verification will be similar to those currently used at the McGuire Nuclear Station in the two-region storage racks installed in both pools. In conclusion, the primary benefit of the McGuire rerack effort has obviously been the amount of storage expansion it provided. A total increase of about 2,000 storage cells was realized, 1,000 of which were the result of pursuing the two-region rather than the conventional poison rack design. Less impacting, but equally as important, however, has been the experience gained during the planning, installation, and operation of these storage racks. This experience should prove useful for future rerack efforts likely to occur at Duke's Catawba Nuclear Station as well as for the current dry storage effort underway for the Oconee Nuclear Station

  6. An independent spent-fuel storage installation at Surry Station: Design and operation

    International Nuclear Information System (INIS)

    McKay, H.S.; Wakeman, B.H.; Pickworth, J.M.; Routh, S.D.; Hopkins, W.C.

    1989-07-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation (ISFSI) was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry ISFSI will accommodate 84 such casks with a total storage capacity of 811 MTU of spent PWR fuel assemblies. The ISFSI is located at the Surry Station in a wooded area approximately 1000 meters (3300 feet) east of the reactor facilities. Construction of the first of three reinforced concrete storage pads and its associated support systems was completed in March 1986. The operating license and Technical Specifications were issued by the US NRC on July 2, 1986. Initial loading operations of a General Nuclear Systems, Inc., CASTOR V/21 storage cask began in September 1986. The first two CASTOR V/21 casks were placed in storage at the ISFSI in December 1986. 16 refs., 33 figs., 16 tabs

  7. In-service inspection of ET-RR-1 reactor vessels and spent fuel storage tank

    International Nuclear Information System (INIS)

    Khattab, M.; Shafy, M.; Konoplev, K.; Samodurve, YU.; Orlov, S.; Didenko, V.; Jackorev, O.

    1993-01-01

    Technical survey included in-service inspection are needed in order to investigate the structural integrity and to insure safe operation of the ET-R R-1 reactor after thirty years aging. An intensive work for the inspection of the inspection of the central tank, shield tank, horizontal channels, primary coolant circuit and spent fuel storage tank have been carried out. The inspection procedures were visual method using video camera and magnification optical as well as thickness measurements using ultrasonic gauge meter and replica for determining defect depth. Water chemical analysis of the primary cooling circuit and spent fuel storage were helpful in results explanation. The results showed that the reactor vessels have good surface conditions. The observed pitting did not affect the structural integrity. The majority of the defects were pits having maximum surface area of about 50 mm. Their depth does not exceed 2 mm. The pits depth rate penetration is of the order of 0.5% per year. Thickness measurements showed insignificant variation. Water status and its chemical properties are very important in controlling corrosion rate. 18 figs., 14 tabs

  8. Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aghoyeh, Reza Gholizadeh [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hosein, E-mail: hkhalafi@aeoi.org.i [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

    2010-10-15

    Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

  9. Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

    International Nuclear Information System (INIS)

    Aghoyeh, Reza Gholizadeh; Khalafi, Hosein

    2010-01-01

    Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

  10. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  11. Design and operational experience of the NUHOMS-24P spent fuel storage system

    International Nuclear Information System (INIS)

    McConaghy, W.J.; Lehnert, R.A.; Rasmussen, R.W.

    1991-01-01

    The NUHOMS spent fuel storage system provides a safe and economical method for the dry storage of spent fuel assemblies either at an independent spent fuel storage installation (ISFSI) at reactor or at a centralized storage facility away from reactor. The system consists of three major safety-related components: a dry shielded canister (DSC) which provides a high integrity containment boundary and a controlled storage environment for the fuel; a reinforced concrete horizontal storage module (HSM) which houses the stored DSCs and provides radiation shielding, protection against natural phenomena and an efficient means for decay heat removal; and a transfer cask which provides for the safe shielded transfer of DSCs from a plant spent fuel pool to a HSM. The NUHOMS system is designed and licensed to the requirements of 10 CFR 72 and ANS/ANSI 57.9 for ISFSIs. The NUHOMS concept was developed in early 1980s, and in 1987, a larger version of the NUHOMS system, 24P, was developed. The operational features of NUHOMS and the loading experience at Oconee are reported. (K.I.)

  12. Status and current spent fuel storage practices in the United States

    International Nuclear Information System (INIS)

    Lake, W.H.

    1999-01-01

    Brief discussions are presented on the history and state of spent fuel generation by utilities that comprise the United States commercial nuclear power industry, the current situation regarding the Federal government's nuclear waste policy, and evolving spent fuel storage practices. These evolving spent fuel storage practices are the result of private sector initiatives, but appear to be influenced by various external factors. The paper is not intended to provide a comprehensive appraisal of the storage initiatives being conducted by the private sector. The focus, instead, is on the Federal government's role and activities related to spent fuel management. Although the Federal government has adopted a policy calling for deep geological disposal of spent fuel, the US Congress has recently begun to consider expanding that policy to include a centralized interim storage facility. In the absence of such an expanded policy, the Department of Energy has performed some preliminary activities that would expedite development of a centralized interim storage facility, if Congress were to enact such a policy. The Department's current activities with regard to developing a centralized interim storage facility, which are consistent with the current policy, are described in the paper. The paper also describes two important technical development activities that have been conducted by the Department of Energy to support improved efficiency in spent fuel management. The Department's activities regarding development of a burnup credit methodology, and a dry transfer system are summarized. (author)

  13. Improving of spent fuel monitoring in condition of Slovak wet interim spent fuel storage facility

    International Nuclear Information System (INIS)

    Miklos, M.; Krsjak, V.; Bozik, M.; Vasina, D.

    2008-01-01

    Monitoring of WWER fuel assemblies condition in Slovakia is presented in the paper. The leak tightness results of fuel assemblies used in Slovak WWER units in last 20 years are analyzed. Good experiences with the 'Sipping system' are described. The Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice was build and put in operation in 1986. Since 1999, leak tests of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP facility with external heating for the precise detection of active specimens. Another system for monitoring of fuel assemblies condition was implemented in December 2006 under the name 'SVYPP-440'. First non-active tests started at February 2007 and are described in the paper. Although those systems seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long (several months). Therefore, a new 'on-line' detection system, based on new sorbent KNiFC-PAN for effective 134 Cs and 137 Cs activity was developed. This sorbent was compared with another type of sorbent NIFSIL and results are presented. The design of this detection system and its possible application in the Slovak wet spent fuel storage facility is discussed. For completeness, the initial results of the new system are also presented. (authors)

  14. Seismic upgrading of the spent fuel storage building at Kozloduy NPP

    International Nuclear Information System (INIS)

    Alexandrov, A.; Borov, V.; Jordanov, M.; Karamanski, T.; Mihaylov, K.

    2001-01-01

    The Spent Fuel Storage Building at Kozloduy NPP site has been analysed for new review level earthquake with 0.2 g peak ground acceleration (compared to the initial design basis earthquake with 0.1 g PGA). The preliminary seismic analysis of the existing building structure using the 5% site specific response spectrum showed the need of seismic structural upgrading. Two upgrading concepts were evaluated on the basis of several factors. The main factor considered was preventing the collapse of the hall structure and the travelling cranes on the fuel storage area during and after a SSE. A three dimensional finite element model was created for the investigation of the seismic response of the existing structure and for the design of the building upgrading. The modelling of the heavy travelling crane and its sub-crane structure was one of the key points. Different configurations of the new upgrading and strengthening structures were investigated. Some interesting conclusions have been drawn from the experience in analysing and upgrading of such a complex industrial structure, comprised of elements with substantial differences in material, rigidity, construction and general behaviour. (author)

  15. Water Clarity Simulant for K East Basin Filtration Testing

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, Andrew J.

    2006-01-20

    This document provides a simulant formulation intended to mimic the behavior of the suspended solids in the K East (KE) Basin fuel storage pool. The simulant will be used to evaluate alternative filtration apparatus to improve Basin water clarity and to possibly replace the existing sandfilter. The simulant was formulated based on the simulant objectives, the key identified parameters important to filtration, the composition and character of the KE Basin suspended sludge particles, and consideration of properties of surrogate materials.

  16. Superstructure optimization of biodiesel production from microalgal biomass

    DEFF Research Database (Denmark)

    Rizwan, Muhammad; Lee, Jay H.; Gani, Rafiqul

    2013-01-01

    In this study, we propose a mixed integer nonlinear programming (MINLP) model for superstructure based optimization of biodiesel production from microalgal biomass. The proposed superstructure includes a number of major processing steps for the production of biodiesel from microalgal biomass...... for the production of biodiesel from microalgae. The proposed methodology is tested by implementing on a specific case study. The MINLP model is implemented and solved in GAMS using a database built in Excel. The results from the optimization are analyzed and their significances are discussed....

  17. Superstructure for high current applications in superconducting linear accelerators

    Science.gov (United States)

    Sekutowicz, Jacek [Elbchaussee, DE; Kneisel, Peter [Williamsburg, VA

    2008-03-18

    A superstructure for accelerating charged particles at relativistic speeds. The superstructure consists of two weakly coupled multi-cell subunits equipped with HOM couplers. A beam pipe connects the subunits and an HOM damper is included at the entrance and the exit of each of the subunits. A coupling device feeds rf power into the subunits. The subunits are constructed of niobium and maintained at cryogenic temperatures. The length of the beam pipe between the subunits is selected to provide synchronism between particles and rf fields in both subunits.

  18. Retrofitting the Superstructure of a Large Passenger Ship Using Composites – A Demonstration

    DEFF Research Database (Denmark)

    Karatzas, Vasileios; Hjørnet, Niels; Berggreen, Christian

    2015-01-01

    In this work, the superstructure of a RoPax ferry has been redesigned using composite materials and the new design has been compared to the existing steel superstructure from a structural perspective. To this end, FE models have been developed and the superstructures have been subjected to loadin...

  19. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-09-01

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  20. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    International Nuclear Information System (INIS)

    Lee, Yoon Hee

    2006-02-01

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  1. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee

    2006-02-15

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  2. Technical, economic and institutional aspects of regional spent fuel storage facilities

    International Nuclear Information System (INIS)

    2005-11-01

    A particular challenge facing countries with small nuclear programmes is the preparation for extended interim storage and then disposal of their spent nuclear fuel. The costs and complications of providing for away-from-reactor storage facilities and/or geological repositories for relatively small amounts of spent fuel may be prohibitively high, motivating interest in regional solutions. This publication addresses the technical, economic and institutional aspects of regional spent fuel storage facilities (RSFSF) and is based on the results of a series of meetings on this topic with participants from IAEA Member States. Topics discussed include safety criteria and standards, safeguards and physical protection, fuel acceptance criteria, long term stability of systems and stored fuel, selection of site, infrastructure aspects, storage technology, licensing, operations, transport, decommissioning, as well as research and development. Furthermore the publication comprises economic, financial and institutional considerations including organizations and legal aspects followed by political and public acceptance and ethical considerations. Approaches and processes for implementation are discussed, as well as the overall benefits and risks of implementing a regional facility. It is illustrated that implementing a RSFSF facility would involve simultaneously addressing a wide range of diverse challenges. The appendix to this report tabulates the numerous issues that have been touched upon in the study. It appears, however, from the discussions that the challenges can in principle be met; the RSFSF concept is technically feasible and potentially economically viable. The technical committees producing this report did not identify any obvious institutional deficiencies that would prevent completion of such a project. Storing spent fuel in a few safe, reliable, secure facilities could enhance safeguards, physical protection and non-proliferation benefits. The committee also

  3. Test plan for K Basin Sludge Canister and Floor Sampling Device

    International Nuclear Information System (INIS)

    Meling, T.A.

    1995-01-01

    This document provides the test plan and procedure forms for conducting the functional and operational acceptance testing of the K Basin Sludge Canister and Floor Sampling Device(s). These samplers samples sludge off the floor of the 100K Basins and out of 100K fuel storage canisters

  4. Superstructure-based optimization of biorefinery networks: Production of biodiesel

    DEFF Research Database (Denmark)

    Bertran, Maria-Ona; Orsi, Albert; Gani, Rafiqul

    2015-01-01

    through a practical case study for the production biodiesel from a variety of feedstock. The different biorefinery processing alternatives are represented in a superstructure and the associated data is collected and stored in a database. Once a specific biorefinery synthesis problem is formulated...

  5. Standardized, utility-DOE compatible, spent fuel storage-transport systems

    International Nuclear Information System (INIS)

    Smith, M.L.

    1991-01-01

    Virginia Power has developed and licensed a facility for dry storage of spent nuclear fuel in metal spent fuel storage casks. The modifications to the design of these casks necessary for licensing for both storage and transport of spent fuel are discussed along with the operational advantages of dual purpose storage-transport casks. Dual purpose casks can be used for storage at utility and DOE sites (MRS or repository) and for shipment between these sites with minimal spent fuel handling. The cost for a standardized system of casks that are compatible for use at both DOE and utility sites is discussed along with possible arrangements for sharing both the cost and benefits of dual purpose storage-transport casks

  6. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    International Nuclear Information System (INIS)

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask's primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives

  7. Postulated licensing schedule for an independent spent-fuel-storage installation

    International Nuclear Information System (INIS)

    Ludwick, J.D.

    1982-11-01

    A review of licensing requirements, processes, and anticipated actions for independent spent fuel storage installations (ISFSIs) was conducted in order to develop an estimated schedule and sequence of events for licensing a new ISFSI. This estimate will be useful to potential ISFSI owners in planning for the licensing of their facilities. It is concluded that, although many uncertainties exist with respect to such things as legal appeals, about 29 months are estimated to elapse between license application and license issuance for an ISFSI. This estimate is in reasonable agreement with a previous time estimate for licensing an ISFSI, and, taking into account the special circumstances involved, with the actual licensing schedule for the GE-Morris ISFSI. However, individual portions of the licensing schedule from each case studied sometimes vary significantly

  8. Simulation of interim spent fuel storage system with discrete event model

    International Nuclear Information System (INIS)

    Yoon, Wan Ki; Song, Ki Chan; Lee, Jae Sol; Park, Hyun Soo

    1989-01-01

    This paper describes dynamic simulation of the spent fuel storage system which is described by statistical discrete event models. It visualizes flow and queue of system over time, assesses the operational performance of the system activities and establishes the system components and streams. It gives information on system organization and operation policy with reference to the design. System was tested and analyzed over a number of critical parameters to establish the optimal system. Workforce schedule and resources with long processing time dominate process. A combination of two workforce shifts a day and two cooling pits gives the optimal solution of storage system. Discrete system simulation is an useful tool to get information on optimal design and operation of the storage system. (Author)

  9. Relative risk measure suitable for comparison of design alternatives of interim spent nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Ferjencik, M.

    1997-01-01

    Accessible reports on risk assessment of interim spent nuclear fuel storage facilities presume that only releases of radioactive substances represent undesired consequences. However, only certain part of the undesired consequences is represented by them. Many other events are connected with safety and are able to cause losses to the operating company. The following two presumptions are pronounced based on this. 1. Any event causing a disturbance of a safety function of the storage facility is an incident event. 2. Any disturbance of a safety function is an undesired consequence. If the facility safety functions are identified and if the severity of their disturbances is quantified, then it is possible to combine consequence severity quantifications and event frequencies into a risk measure. Construction and application of such a risk measure is described in this paper. The measure is shown to be a tool suitable for comparison of interim storage technology design alternatives. (author)

  10. Spent fuel storage requirements for nuclear utilities and OCRWM [Office of Civilian Radioactive Waste Management

    International Nuclear Information System (INIS)

    Wood, T.W.

    1990-03-01

    Projected spent fuel generation at US power reactors exceeds estimated aggregate pool storage capacity by approximately 30,000 metric tons of uranium (MTU). Based on the current repository schedule, little of the spent fuel inventory will be disposed of prior to shutdown of existing reactors, and a large additional capacity for surface storage of spent fuel will be required, either at reactors or at a centralized DOE storage site. Allocation of this storage requirement across the utility-DOE interface, and the resulting implications for reactor sites and the performance of the federal waste management system, were studied during the DOE MRS System Study and again subsequent to the reassessment of the repository schedule. Spent fuel logistics and cost results from these analyses will be used in definition of spent fuel storage capacity requirements for the federal system. 9 refs., 8 figs., 1 tab

  11. Partition wall structure in spent fuel storage pool and construction method for the partition wall

    International Nuclear Information System (INIS)

    Izawa, Masaaki

    1998-01-01

    A partitioning wall for forming cask pits as radiation shielding regions by partitioning inside of a spent fuel storage pool is prepared by covering both surface of a concrete body by shielding metal plates. The metal plate comprises opposed plate units integrated by welding while sandwiching a metal frame as a reinforcing material for the concrete body, the lower end of the units is connected to a floor of a pool by fastening members, and concrete is set while using the metal plate of the units as a frame to form the concrete body. The shielding metal plate has a double walled structure formed by welding a lining plate disposed on the outer surface of the partition wall and a shield plate disposed to the inner side. Then the term for construction can be shortened, and the capacity for storing spent fuels can be increased. (N.H.)

  12. Study on uncertainty evaluation system for the safety evaluation of interim spent fuel storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyeon; Shin, Myeong Won; Rhy, Seok Jin; Cho, Dong Keon; Park, Dong Hwan [Kyunghee Univ., Seoul (Korea, Republic of); Cheong, Beom Jin [Minstry of Science and Technology, Gwacheon (Korea, Republic of)

    1998-03-15

    The main objective os to develop a technical standards for the facility operation of the interm, spent fuel storage facility and to develop a draft for the technical criteria to be legislated. The another objective os to define a uncertainty evaluation system for burn up credit application in criticality analysis and to investigate an applicability of this topic for future regulatory activity. Investigate a status of art for the operational criteria of spent fuel interm wet storage. Collect relevant laws, decree, notices and standards related to the operation of storage facility and study on the legislation system. Develop a draft of technical standards and criteria to be legislated. Define an evaluation system for the uncertainty analysis and study on the status of art in the field of criticality safety analysis. Develop an uncertainty evaluation system in criticality analysis with burnup credit and investigate an applicability as well as its benefits of this policy.

  13. Cost update: Technology, safety, and costs of decommissioning reference independent spent fuel storage installations

    International Nuclear Information System (INIS)

    Miles, T.L.

    1994-07-01

    The cost estimates originally developed in NUREG/CR-2210 for decommissioning five conceptual Independent Spent Fuel Storage Installations (ISFSIs) and their supporting ancillaries (hot cell and transporter) are updated from 1981 to 1993 dollars. The costs for labor and materials increased approximately at the rate of inflation, the cost of energy increased more slowly than the rate of inflation, and the cost of low-level radioactive waste disposal increased much more rapidly than the rate of inflation. A methodology and a formula are presented for estimating the cost of decommissioning the ISFSIs at some future time, based on these current cost estimates. The formula contains essentially the same elements as the formula given in 10 CFR 50.75 for escalating the decommissioning costs for nuclear power reactors to some future time

  14. Changing the Rules on Fuel Export at Sellafield's First Fuel Storage Pond - 12065

    Energy Technology Data Exchange (ETDEWEB)

    Carlisle, Derek [Sellafield Ltd, Sellafield, Cumbria (United Kingdom)

    2012-07-01

    The Pile Fuel Storage Pond (PFSP) was built in 1949/50 to receive, store and de-can fuel and isotopes from the Windscale Piles. Following closure of the Piles in 1957, plant operations were scaled down until fuel processing eventually ceased in 1962. The facility has held an inventory of metal fuel both from the Piles and from other programmes since that time. The pond is currently undergoing remediation and removal of the fuel is a key step in that process, unfortunately the fuel export infrastructure on the plant is no longer functional and due to the size and limited lifting capability, the plant is not compatible with today's large volume heavy export flasks. The baseline scheme for the plant is to package fuel into a small capacity flask and transfer it to another facility for treatment and repackaging into a flask compatible with other facilities on site. Due to programme priorities the repackaging facility is not available to do this work for several years causing a delay to the work. In an effort accelerate the programme the Metal Fuel Pilot Project (MFPP) was initiated to challenge the norms for fuel transfer and develop a new methodology for transferring the fuel. In developing a transfer scheme the team had to overcome challenges associated with unknown fuel condition, transfers outside of bulk containment, pyro-phoricity and oxidisation hazards as well as developing remote control and recovery systems for equipment not designed for this purpose. A combination of novel engineering and enhanced operational controls were developed which resulted in the successful export of the first fuel to leave the Pile Fuel Storage Pond in over 40 years. The learning from the pilot project is now being considered by the main project team to see how the new methodology can be applied to the full inventory of the pond. (author)

  15. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  16. Analysis of ICPP fuel storage rack inner tie and corner tie substructures

    Energy Technology Data Exchange (ETDEWEB)

    Nitzel, M.E.; Rahl, R.G.

    1996-01-01

    Finite element models were developed and analyses performed for the tie plate, inner tie block assembly, and corner tie block assembly of a 25 port fuel rack assembly designed for installation in Pool 1 of Building 666 at the Idaho Chemical Processing Plant. These models were specifically developed to investigate the adequacy of certain welds joining components of the fuel storage rack assembly. The work scope for the task was limited to an investigation of the stress levels in the subject subassemblies when subjected to seismic loads. Structural acceptance criteria used for the elastic calculations performed were as found in the overall rack design report as issued by the rack`s designer, Holtec International. Structural acceptance criteria used for the plastic calculations performed as part of this effort were as defined in Subsection NF and Appendix F of the ASME Boiler & Pressure Vessel Code. The results of the analyses will also apply to the 30 port fuel storage rack design that is also scheduled for installation in Pool 1 of ICPP 666. The results obtained from the analyses performed for this task indicate that the welds joining the inner tie block and corner tie block to the surrounding rack structure meet the acceptance criteria. Further, the structural members (plates and blocks) were also found to be within the allowable stress limits established by the acceptance criteria. The separate analysis performed on the inner tie plate confirmed the structural adequacy for both the inner tie plate, corner tie plate, and tie block bolts. The analysis results verified that the inner tie and corner tie block should be capable of transferring the expected seismic load without structural failure.

  17. Seismic justification of free standing spent fuel storage racks: experimental versus computed behavior

    International Nuclear Information System (INIS)

    Champomier, F.P.; Peron, J.Y.

    2001-01-01

    Most spent fuel storage racks now designed for nuclear power plants are of the free standing type. They consist in a series of independent modules of varying size (typically, 2 m long and wide and 5 m high) immersed in water but simply left standing on the bottom of the spent fuel storage pool. The ever increasing need for additional storage room leads power plants to wish to have compact modules positioned as close to each other as possible. The reduced clearance is essentially limited by the possible interaction of a module with its neighbors in case of a seismic event. In fact, during an earthquake, it is anticipated that each module may have a complex and different movement relative to the pool floor: sliding, uplifting, rocking may occur depending on various parameters related to module geometry and load and seismic excitation. Minimum clearance between modules is to be defined and justified. Two approaches are possible: each module may be demonstrated to stay wide from its neighbors or to impact them with no significant resulting damage. Though experimental work is sometimes possible, it certainly cannot be routinely or systematically envisaged. Thus, approaches to evaluation and justification basically rely on numerical simulation of the problem. Actually, the entire calculation scheme is often repeated a large number of times with varying parameters (seismic input in the three directions, friction coefficients, fuel loading characteristics, etc.) for additional treatment and confidence. The object of this paper is to provide some insight on the kind of information one can reasonably expect to obtain from such studies. (author)

  18. Spent-fuel-storage requirements: an update of DOE/RL-82-1

    International Nuclear Information System (INIS)

    1983-01-01

    Spent fuel storage requirements as projected through the year 2000 for US light water reactor (LWR) nuclear power plants were calculated using information supplied by the utilities reflecting plant status as of September 30, 1982. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 132 gigawatts electrical (GWe) projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities, as estimated by the utilities, for spent fuel storage utilizing currently licensed technologies. Rod consolidation and dry storage technologies were not considered. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 metric tons uranium (MTU) of additional storage in 1984, and a cumulative requirement for 13,090 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR) within a given utility. In all cases, a full core discharge capability is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. 1 figure, 12 tables

  19. Studies and research concerning BNFP: life of project operating expenses for away-from-reactor (AFR) spent fuel storage facility. Final report

    International Nuclear Information System (INIS)

    Shallo, F.A.

    1979-09-01

    Life of Project operating expenses for a licensed Away-From-Reactor (AFR) Spent Fuel Storage Facility are developed in this report. A comprehensive business management structure is established and the functions and responsibilities for the facility organization are described. Contractual provisions for spent fuel storage services are evaluated

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

    Science.gov (United States)

    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...

  1. 78 FR 22411 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Amendment No. 8; Corrections

    Science.gov (United States)

    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...

  2. 76 FR 12825 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Confirmation of...

    Science.gov (United States)

    2011-03-09

    ... Storage Casks: NUHOMS[supreg] HD System Revision 1; Confirmation of Effective Date AGENCY: Nuclear... direct final rule amended the NRC's spent fuel storage regulations at Title 10 of the Code of Federal Regulations (10 CFR 72.214) to revise the NUHOMS[supreg] HD System listing to include Amendment Number 1 to...

  3. Corrective Action Investigation Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada; TOPICAL

    International Nuclear Information System (INIS)

    1999-01-01

    This Corrective Action Investigation Plan (CAIP) has been developed in accordance with the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the US Department of Energy, Nevada Operations Office (DOE/NV); the State of Nevada Division of Environmental Protection (NDEP); and the US Department of Defense (FFACO, 1996). The CAIP is a document that provides or references all of the specific information for investigation activities associated with Corrective Action Units (CAUs) or Corrective Action Sites (CASs). According to the FFACO (1996), CASs are sites potentially requiring corrective action(s) and may include solid waste management units or individual disposal or release sites. A CAU consists of one or more CASs grouped together based on geography, technical similarity, or agency responsibility for the purpose of determining corrective actions. This CAIP contains the environmental sample collection objectives and the criteria for conducting site investigation activities at the CAU 321 Area 22 Weather Station Fuel Storage, CAS 22-99-05 Fuel Storage Area. For purposes of this discussion, this site will be referred to as either CAU 321 or the Fuel Storage Area. The Fuel Storage Area is located in Area 22 of the Nevada Test Site (NTS). The NTS is approximately 105 kilometers (km) (65 miles[mi]) northwest of Las Vegas, Nevada (Figure 1-1) (DOE/NV, 1996a). The Fuel Storage Area (Figure 1-2) was used to store fuel and other petroleum products necessary for motorized operations at the historic Camp Desert Rock facility which was operational from 1951 to 1958 at the Nevada Test Site, Nevada. The site was dismantled after 1958 (DOE/NV, 1996a)

  4. Laboratory Evaluation of Underwater Grouting of CPP-603 Basins

    International Nuclear Information System (INIS)

    Johnson, V.J.; Pao, J.H.; Demmer, R.L.; Tripp, J.L.

    2002-01-01

    A project is underway to deactivate a Fuel Storage Basin. The project specifies the requirements and identifies the tasks that will be performed for deactivation of the CPP- 603 building at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory. The Fuel Receiving and Storage Building (CPP- 603) was originally used to receive and store spent nuclear fuel from various facilities. The area to undergo deactivation includes the three spent nuclear fuel storage basins and a transfer canal (1.5 million gallons of water storage). Deactivation operations at the task site include management of the hot storage boxes and generic fuel objects, removal of the fuel storage racks, basin sludge, water evaporation and basin grouting, and interior equipment, tanks, and associated components. This includes a study to develop a grout formulation and placement process for this deactivation project. Water will be allowed to passively evaporate to r educe the spread of contamination from the walls of the basin. The basins will be filled with grout, underwater, as the water evaporates to maintain the basin water at a safe level. The objective of the deactivation project is to eliminate potential exposure to hazardous and radioactive materials and eliminate potential safety hazards associated with the CPP-603 building

  5. SELF-ORGANIZATION OF LEAD SULFIDE QUANTUM DOTS INTO SUPERSTRUCTURES

    Directory of Open Access Journals (Sweden)

    Elena V. Ushakova

    2014-11-01

    Full Text Available The method of X-ray structural analysis (X-ray scattering at small angles is used to show that the structures obtained by self-organization on a substrate of lead sulfide (PbS quantum dots are ordered arrays. Self-organization of quantum dots occurs at slow evaporation of solvent from a cuvette. The cuvette is a thin layer of mica with teflon ring on it. The positions of peaks in SAXS pattern are used to calculate crystal lattice of obtained ordered structures. Such structures have a primitive orthorhombic crystal lattice. Calculated lattice parameters are: a = 21,1 (nm; b = 36,2 (nm; c = 62,5 (nm. Dimensions of structures are tens of micrometers. The spectral properties of PbS QDs superstructures and kinetic parameters of their luminescence are investigated. Absorption band of superstructures is broadened as compared to the absorption band of the quantum dots in solution; the luminescence band is slightly shifted to the red region of the spectrum, while its bandwidth is not changed much. Luminescence lifetime of obtained structures has been significantly decreased in comparison with the isolated quantum dots in solution, but remained the same for the lead sulfide quantum dots close-packed ensembles. Such superstructures can be used to produce solar cells with improved characteristics.

  6. Mach Number effects on turbulent superstructures in wall bounded flows

    Science.gov (United States)

    Kaehler, Christian J.; Bross, Matthew; Scharnowski, Sven

    2017-11-01

    Planer and three-dimensional flow field measurements along a flat plat boundary layer in the Trisonic Wind Tunnel Munich (TWM) are examined with the aim to characterize the scaling, spatial organization, and topology of large scale turbulent superstructures in compressible flow. This facility is ideal for this investigation as the ratio of boundary layer thickness to test section spanwise extent ratio is around 1/25, ensuring minimal sidewall and corner effects on turbulent structures in the center of the test section. A major difficulty in the experimental investigation of large scale features is the mutual size of the superstructures which can extend over many boundary layer thicknesses. Using multiple PIV systems, it was possible to capture the full spatial extent of large-scale structures over a range of Mach numbers from Ma = 0.3 - 3. To calculate the average large-scale structure length and spacing, the acquired vector fields were analyzed by statistical multi-point methods that show large scale structures with a correlation length of around 10 boundary layer thicknesses over the range of Mach numbers investigated. Furthermore, the average spacing between high and low momentum structures is on the order of a boundary layer thicknesses. This work is supported by the Priority Programme SPP 1881 Turbulent Superstructures of the Deutsche Forschungsgemeinschaft.

  7. Conceptual design study of a concrete canister spent-fuel storage facility

    International Nuclear Information System (INIS)

    Lidfors, E.D.; Tabe, T.; Johnson, H.M.

    1979-01-01

    This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

  8. Transmutation technologies to solve the problem of long-term spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Hosnedl, P.; Valenta, V.; Blahut, O.

    2000-01-01

    The paper gives a brief description of the transmutation process for actinides and long-lived fission products which are present in spent nuclear fuel. Transmutation technologies can solve the problem of long-term spent nuclear fuel storage and reduce the requirements for storage time and conditions. The basic data and requirements for the detailed design of the transmutor are summarized, and the views upon how to address the fuel purification and dry reprocessing issues are discussed. The results of activities of SKODA JS are highlighted; these include, for instance, the fluoride salt-resistant material MONICR, test loops, and electrowinners. The preliminary design of the transmutor is also outlined. Brief information regarding activities in the field of transmutation technologies in the Czech Republic and worldwide is also presented. The research and design activities to be developed for the whole design of the demonstration and basic units are summarized. It is emphasized that SKODA JS can join in international cooperation without constraints. The Attachment presents a simple assessment of how the radioactivity balance can be reduced, based on the actinide and long-lived fission product transmutation half-lives, is presented in the Attachment. (author)

  9. Test on Similarity between the Flooded and Optimum Moderation Conditions of the Spent Fuel Storage Pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gil Soo; Jang, Chang Sun; Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    In the criticality safety analysis, uncertainty and bias should be considered. The final multiplication factor including uncertainty and bias in addition to calculated k-eff should be below the administrative limit. The administrative limit of spent fuel pool is 0.95 with flooded condition (filled with unborated water), and 0.98 with optimum moderation condition (filled with foggy unborated water, usually occurs near 0.1g/cc water density) for new fuel storage. The bias is determined by comparing the calculation results of the critical experiments ever performed. It is important to choose 'good' experiments which have 'similar' condition with application. To obtain realistic bias, many experiments with similar conditions should be chosen and considered. In previous approach, same critical experiment set are used to determine bias of the flooded and optimum moderation conditions. It would be correct way if two conditions are similar. The similarity test on this paper was performed by TSUNAMI code included in SCALE5.1 package. TSUNAMI code produces sensitivity data for each nuclear reaction by using first order perturbation theory. TSUNAMI code performs forward and adjoint multigroup Monte Carlo calculation. Sensitivity data are obtained by forward and adjoint results. TSUNAMI also produces uncertainty data with sensitivity data and cross section covariance data. In this paper, similarity is determined by comparing energy of average lethargy of fission (EALF), uncertainty data, sensitivity data, and correlation coefficient which is also output of the TSUNAMI code.

  10. CASTOR-V/21 PWR spent fuel storage cask performance test

    International Nuclear Information System (INIS)

    Creer, J.M.; Schoonen, D.H.

    1986-01-01

    Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

  11. Risk Analysis of a Fuel Storage Terminal Using HAZOP and FTA

    Directory of Open Access Journals (Sweden)

    José Luis Fuentes-Bargues

    2017-06-01

    Full Text Available The size and complexity of industrial chemical plants, together with the nature of the products handled, means that an analysis and control of the risks involved is required. This paper presents a methodology for risk analysis in chemical and allied industries that is based on a combination of HAZard and OPerability analysis (HAZOP and a quantitative analysis of the most relevant risks through the development of fault trees, fault tree analysis (FTA. Results from FTA allow prioritizing the preventive and corrective measures to minimize the probability of failure. An analysis of a case study is performed; it consists in the terminal for unloading chemical and petroleum products, and the fuel storage facilities of two companies, in the port of Valencia (Spain. HAZOP analysis shows that loading and unloading areas are the most sensitive areas of the plant and where the most significant danger is a fuel spill. FTA analysis indicates that the most likely event is a fuel spill in tank truck loading area. A sensitivity analysis from the FTA results show the importance of the human factor in all sequences of the possible accidents, so it should be mandatory to improve the training of the staff of the plants.

  12. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

  13. Development of maintenance technology with underwater TIG welding for spent fuel storage pool

    International Nuclear Information System (INIS)

    Obana, Takeshi; Hamada, Yasumitsu; Ooeda, Kaoru; Katou, Masahide; Ootsuka, Toshihiro; Toyoda, Seiichi; Hosogane, Atsushi

    2007-01-01

    The core technology of underwater TIG welding process has been developed and welding equipment system has been manufactured, for application to the maintenance of the spent fuel storage pool of Rokkasho reprocessing plant. Basic experiments for understanding the conditions of dry area and the range of welding conditions was performed, and mock examination for simulation of real environment by using the developed welding equipment was also carried out to judge the applicability of the system. For the purpose that can be selected water removing method for different spatial conditions of the parts to be maintained in underwater, two kinds of welding equipment systems of Chamber type and Partition type were developed and manufactured. On the basis of fundamental experiments, the conditions of dry area formation and welding parameters range for high-reliability weld were discussed. Thus the proper condition in this process was able to be established. With the welding equipment systems of the Chamber type and Partition type, the practical use examination of underwater TIG welding process was executed by mock examination for simulating the real environment. As a result, it was confirmed that the underwater TIG welding could obtain the same reliability as a usual in-air TIG welding, and the operation and the control at remote distance were also possible. And the reliability of the patch-plate fillet weld could be evaluated by remote inspection with the expansion visual test. (author)

  14. Risk Analysis of a Fuel Storage Terminal Using HAZOP and FTA.

    Science.gov (United States)

    Fuentes-Bargues, José Luis; González-Cruz, Mª Carmen; González-Gaya, Cristina; Baixauli-Pérez, Mª Piedad

    2017-06-30

    The size and complexity of industrial chemical plants, together with the nature of the products handled, means that an analysis and control of the risks involved is required. This paper presents a methodology for risk analysis in chemical and allied industries that is based on a combination of HAZard and OPerability analysis (HAZOP) and a quantitative analysis of the most relevant risks through the development of fault trees, fault tree analysis (FTA). Results from FTA allow prioritizing the preventive and corrective measures to minimize the probability of failure. An analysis of a case study is performed; it consists in the terminal for unloading chemical and petroleum products, and the fuel storage facilities of two companies, in the port of Valencia (Spain). HAZOP analysis shows that loading and unloading areas are the most sensitive areas of the plant and where the most significant danger is a fuel spill. FTA analysis indicates that the most likely event is a fuel spill in tank truck loading area. A sensitivity analysis from the FTA results show the importance of the human factor in all sequences of the possible accidents, so it should be mandatory to improve the training of the staff of the plants.

  15. Improved method to demonstrate the structural integrity of high density fuel storage racks

    International Nuclear Information System (INIS)

    Hinderks, M.; Ungoreit, H.; Kremer, G.

    2001-01-01

    Reracking of existing fuel pools to the maximum extent is desirable from an economical point of view. This goal can be achieved by minimizing the gaps between the spent fuel storage racks. Since the rack design is aimed at enabling consolidated fuel rod storage, additional requirements arise with respect to the design and the structural analysis. The loads resulting from seismic events are decisive for the structural analysis and require a specially detailed and in-depth analysis for high seismic loads. The verification of structural integrity and functionality is performed in two phases. In the first phase the motional behavior of single racks, rows of racks and, where required, of all racks in the pool is simulated by excitation with displacement time histories under consideration of the fluid-structure interaction (FSI). The displacements from these simulations are evaluated, while the loads are utilized as input data for the structural analysis of the racks and the pool floor. The structural analyses for the racks comprise substantially stress analyses for base material and welds as well as stability analyses for the support channels and the rack outside walls. The analyses are performed in accordance with the specified codes and standards

  16. Optimization strategies for cask design and container loading in long term spent fuel storage

    International Nuclear Information System (INIS)

    2006-12-01

    As delays are incurred in implementing reprocessing and in planning for geologic repositories, storage of increasing quantities of spent fuel for extended durations is becoming a growing reality. Accordingly, effective management of spent fuel continues to be a priority topic. In response, the IAEA has organized a series of meetings to identify cask loading optimisation issues in preparation for a technical publication on Optimization Strategies for Cask/Container Loading in Long Term Spent Fuel Storage. This publication outlines the optimisation process for cask design, licensing and utilization, describing three principal groups of optimization activities in terms of relevant technical considerations such as criticality, shielding, structural design, operations, maintenance and retrievability. The optimization process for cask design, licensing, and utilization is outlined. The general objectives for the design of storage casks, including storage casks that are intended to be transportable, are summarized. The nature of optimization within the design process is described. The typical regulatory and licensing process is outlined, focusing on the roles of safety regulations, the regulator, and the designer/applicant in the optimization process. Based on the foregoing, a description of the three principal groups of optimization activities is provided. The subsequent chapters of this document then describe the specific optimization activities within these three activity groups, in each of the several design disciplines

  17. Preliminary study on detection technology of the cladding weld of spent fuel storage pool

    Science.gov (United States)

    Qi, Pan; Cui, Hongyan; Feng, Meiming; Shao, Wenbin; Liao, Shusheng; Li, Wei

    2018-04-01

    As the first barrier of the Spent fuel storage pool, the steel cladding using different sizes (length×width) of 304L stainless steel with 3˜6mm thickness plate argon arc welded together which is direct contacted with boric acid water. Environmental humidity between the back of steel cladding and concrete, makes phosphate, chloride ion overflowed from the concrete that corroded on the weld zone with different mechanism. Part of the corrosion defects can penetrate leaded to leakage of boric acid water in penetration position accelerated crack propagation. In view of the above situation and combined with the actual needs of the power plant, the development of effective underwater nondestructive testing means of the weld area for periodic inspection and monitoring is necessary. A single method may lead to the missing of defects detection due to weld reinforcement unpolished. In this paper, eddy current array (ARRAY) and Alternating Current Field Measurement (ACFM) are adapted to test the limit sensitivity and resolution through by the specimens with artificial defects which make their detection abilities close to satisfy engineering requirements. The preliminary study found that Φ0.5mm through-wall hole and with 2mm length and 0.3mm width through-wall crack in the weld can be good inspected.

  18. Welding issues associated with design, fabrication and loading of spent fuel storage casks

    International Nuclear Information System (INIS)

    Battige, C.K. Jr.; Howe, A.G.; Sturz, F.C.

    1999-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has observed a number of welding issues associated with design, fabrication, and loading of spent fuel storage casks. These emerging welding-related issues involving a certain dry cask storage system have challenged the safety basis for which NRC approved the casks for storage of spent nuclear fuel. During closure welding, problems have been encountered with cracking. Although the cracks have been attributed to several causes including material suitability, joint restraint and residual stresses, NRC believes hydrogen-induced cracking is the most likely explanation. In light of these cracking events and the potential for flaws in any welding process, NRC sought verification of the corrective actions and the integrity of the lid closure welds before allowing additional casks to be loaded. As a result, the affected utility companies modified the closure welding procedures and developed an acceptable ultrasonic inspection (UT) method. In addition, the casks already loaded at three power reactor sites will require additional non-destructive examinations (NDE) to determine their suitability for continued use. NRC plans to evaluate the generic implications of this issue for current designs and for those in the licensing process. (author)

  19. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Eslinger, L.E.; Schmitt, R.C.

    1989-01-01

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

  20. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

  1. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

  2. An Evaluation of Criticality Margin by an Application of Parallelogram Lattice Arrangement in the Nuclear Fuel Storage Rack

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Kim, Hong Chul; Shin, Chang Ho; Kim, Jong Kyung; Kim, Kyo Youn

    2010-01-01

    The criticality evaluation in the nuclear fuel storage rack is essentially required for the prevention of the criticality accident. The square lattice structure of the storage rack is commonly used because it has a simple structure for the storage of the numerous fuel assemblies as well as the good mechanical strength. For the design of the fuel storage rack, the boron plate is commonly used for the criticality reduction. In this study, an arrangement method with the parallelogram lattice structure is proposed for the reduction of the boron concentration or the rack pitch. The criticality margins by the application of the parallelogram lattice were evaluated with MCNP5 code. From the result, the reduction of the boron concentrated in the borated-Al plate was evaluated

  3. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation

    International Nuclear Information System (INIS)

    Hall, Gregory Graham

    2002-01-01

    This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  4. Castor-V/21 PWR spent fuel storage cask performance test

    International Nuclear Information System (INIS)

    Creer, J.M.; Schoonen, D.H.

    1986-01-01

    Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

  5. CPP-603 Underwater Fuel Storage Facility Site Integrated Stabilization Management Plan (SISMP), Volume I

    International Nuclear Information System (INIS)

    Denney, R.D.

    1995-10-01

    The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been constructed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1 to the Secretary of Energy regarding an improved schedule for remediation in the Defense Nuclear Facilities Complex. As described in the DNFSB Recommendation 94-1 Implementation Plan, issued February 28, 1995, an INEL Spent Nuclear Fuel Management Plan is currently under development to direct the placement of SNF currently in existing INEL facilities into interim storage, and to address the coordination of intrasite SNF movements with new receipts and intersite transfers that were identified in the DOE SNF Programmatic and INEL Environmental Restoration and Waste Management Environmental Impact Statement Record, of Decision. This SISMP will be a subset of the INEL Spent Nuclear Fuel Management Plan and the activities described are being coordinated with other INEL SNF management activities. The CPP-603 relocation activities have been assigned a high priority so that established milestones will be meet, but there will be some cases where other activities will take precedence in utilization of available resources. The Draft INEL Site Integrated Stabilization Management Plan (SISMP), INEL-94/0279, Draft Rev. 2, dated March 10, 1995, is being superseded by the INEL Spent Nuclear Fuel Management Plan and this CPP-603 specific SISMP

  6. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  7. Criticality analysis of the Annular Core Pulse Reactor (ACPR) fuel storage container

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J S [Sandia Laboratories (United States)

    1974-07-01

    The ACPR fuel storage rack is a water moderated steel frame assembly with aluminum guide tubes and grid plates. The rack has a capacity for 90 fuel elements - 10 rows of 9 elements each. A section, four inches wide, in the center of the rack is reserved for a neutron source and detectors. Quarter-inch boral plate separates each row of fuel elements from its adjacent row(s). The storage rack was analyzed by generating cell-disadvantaged cross sections for the fuel element rows so that the rows could be treated as homogeneous regions in slab geometry. The rack could then be described in one dimension as a series of parallel slab regions with buckling corrections for the uniform width and height of the rows. The DTF-4 code (S{sub N} transport theory) and 16 energy group cross sections were used for the neutron transport calculations yielding a multiplication factor k{sub eff} = 0.446. Further calculations were performed on the fully loaded storage array to assess its subcriticality on the basis of geometry alone, i.e., without taking credit for any burnable or removable poisons such as the boral plates. For these calculations the boral plates were replaced with water and the multiplication factor increased markedly, k{sub eff} = 0.945. Criticality guides (e.g., ANSI N16.5, February 1973) indicate that computed neutron multiplication factors for storage arrays should be <0.95 using validated computational techniques. To demonstrate conclusively that the 0.95 limit is satisfied on the basis of geometry alone, additional calculations (e.g., three dimensional Monte Carlo) or experimental verification may be necessary since there has been no attempt to estimate the error introduced by the one- dimensional model or the cross section. (author)

  8. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    1995-01-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ''fresh fuel assumption'' increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible

  9. Preliminary estimates of the charge for spent-fuel storage and disposal services

    International Nuclear Information System (INIS)

    1978-07-01

    The purpose of this report is to stimulate discussion among a wide range of interested parties concerning a one-time charge by the U.S. Government for disposal, or interim storage and disposal, of spent unreprocessed nuclear fuel. The report contains a set of estimates of the charge based on current cost figures and a variety of demand, logistical, institutional, and cost overrun assumptions. The services are to be offered to domestic utilities by the U.S. Government in connection with the spent fuel policy approved by the President and announced by the Department of Energy (DOE) on October 18, 1977. This policy is a direct result of the indefinite deferral of all commercial reprocessing of spent fuel announced by President Carter on April 7, 1977. The services will also be offered to foreign governments on a limited basis in cases where this action would contribute to U.S. goals for nonproliferation of nuclear weapons. The report does not establish new policy and it does not commit DOE to any specific program, schedule or charge. No scenario or case is to be considered most important, no methodology is to be considered definitive, and no charge is to be considered most likely or to represent a proposed charge. The report describes basic principles and methodologies for calculating the charge and highlights primary cost centers. Current estimates of program and facility costs are used. Various aspects of the DOE Spent Fuel Storage Program are brought into focus through this analysis. Interested parties should find these assessment criteria helpful for their planning and useful in discussions concerning the program

  10. Criticality analysis of the Annular Core Pulse Reactor (ACPR) fuel storage container

    International Nuclear Information System (INIS)

    Philbin, J.S.

    1974-01-01

    The ACPR fuel storage rack is a water moderated steel frame assembly with aluminum guide tubes and grid plates. The rack has a capacity for 90 fuel elements - 10 rows of 9 elements each. A section, four inches wide, in the center of the rack is reserved for a neutron source and detectors. Quarter-inch boral plate separates each row of fuel elements from its adjacent row(s). The storage rack was analyzed by generating cell-disadvantaged cross sections for the fuel element rows so that the rows could be treated as homogeneous regions in slab geometry. The rack could then be described in one dimension as a series of parallel slab regions with buckling corrections for the uniform width and height of the rows. The DTF-4 code (S N transport theory) and 16 energy group cross sections were used for the neutron transport calculations yielding a multiplication factor k eff = 0.446. Further calculations were performed on the fully loaded storage array to assess its subcriticality on the basis of geometry alone, i.e., without taking credit for any burnable or removable poisons such as the boral plates. For these calculations the boral plates were replaced with water and the multiplication factor increased markedly, k eff = 0.945. Criticality guides (e.g., ANSI N16.5, February 1973) indicate that computed neutron multiplication factors for storage arrays should be <0.95 using validated computational techniques. To demonstrate conclusively that the 0.95 limit is satisfied on the basis of geometry alone, additional calculations (e.g., three dimensional Monte Carlo) or experimental verification may be necessary since there has been no attempt to estimate the error introduced by the one- dimensional model or the cross section. (author)

  11. An integrated methodology to evaluate a spent nuclear fuel storage system

    International Nuclear Information System (INIS)

    Yoon, Jeong Hyoun

    2008-02-01

    This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

  12. A Generic Methodology for Superstructure Optimization of Different Processing Networks

    DEFF Research Database (Denmark)

    Bertran, Maria-Ona; Frauzem, Rebecca; Zhang, Lei

    2016-01-01

    In this paper, we propose a generic computer-aided methodology for synthesis of different processing networks using superstructure optimization. The methodology can handle different network optimization problems of various application fields. It integrates databases with a common data architecture......, a generic model to represent the processing steps, and appropriate optimization tools. A special software interface has been created to automate the steps in the methodology workflow, allow the transfer of data between tools and obtain the mathematical representation of the problem as required...

  13. 40 CFR 745.228 - Accreditation of training programs: public and commercial buildings, bridges and superstructures...

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Accreditation of training programs: public and commercial buildings, bridges and superstructures. [Reserved] 745.228 Section 745.228... Accreditation of training programs: public and commercial buildings, bridges and superstructures. [Reserved] ...

  14. 46 CFR 108.133 - Hull superstructure, structural bulkheads, decks, and deckhouses.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Hull superstructure, structural bulkheads, decks, and deckhouses. 108.133 Section 108.133 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) A... Protection § 108.133 Hull superstructure, structural bulkheads, decks, and deckhouses. Each hull...

  15. Role of Steric Hindrance in the Crystal Packing of Z′ = 4 Superstructure of Trimethyltin Hydroxide

    KAUST Repository

    Dey, S.; Schö nleber, A.; Mondal, S.; Ali, S. I.; van Smaalen, S.

    2018-01-01

    The roomerature crystal structure of trimethyltin hydroxide, (CH)SnOH, has been described by Anderson et al. [Cryst. Growth Des. 2011, 11, 820-826] as a 2a × 2b × 8c, 32-fold superstructure. We report a a × b × 8c, eight-fold superstructure

  16. Voids and superstructures: correlations and induced large-scale velocity flows

    Science.gov (United States)

    Lares, Marcelo; Luparello, Heliana E.; Maldonado, Victoria; Ruiz, Andrés N.; Paz, Dante J.; Ceccarelli, Laura; Garcia Lambas, Diego

    2017-09-01

    The expanding complex pattern of filaments, walls and voids build the evolving cosmic web with material flowing from underdense on to high density regions. Here, we explore the dynamical behaviour of voids and galaxies in void shells relative to neighbouring overdense superstructures, using the Millenium simulation and the main galaxy catalogue in Sloan Digital Sky Survey data. We define a correlation measure to estimate the tendency of voids to be located at a given distance from a superstructure. We find voids-in-clouds (S-types) preferentially located closer to superstructures than voids-in-voids (R-types) although we obtain that voids within ˜40 h-1 Mpc of superstructures are infalling in a similar fashion independently of void type. Galaxies residing in void shells show infall towards the closest superstructure, along with the void global motion, with a differential velocity component depending on their relative position in the shell with respect to the direction to the superstructure. This effect is produced by void expansion and therefore is stronger for R-types. We also find that galaxies in void shells facing the superstructure flow towards the overdensities faster than galaxies elsewhere at the same relative distance to the superstructure. The results obtained for the simulation are also reproduced for the Sky Survey Data Release data with a linearized velocity field implementation.

  17. Turbulent thermal superstructures in Rayleigh-Bénard convection

    Science.gov (United States)

    Stevens, Richard J. A. M.; Blass, Alexander; Zhu, Xiaojue; Verzicco, Roberto; Lohse, Detlef

    2018-04-01

    We report the observation of superstructures, i.e., very large-scale and long living coherent structures in highly turbulent Rayleigh-Bénard convection up to Rayleigh Ra=109 . We perform direct numerical simulations in horizontally periodic domains with aspect ratios up to Γ =128 . In the considered Ra number regime the thermal superstructures have a horizontal extend of six to seven times the height of the domain and their size is independent of Ra. Many laboratory experiments and numerical simulations have focused on small aspect ratio cells in order to achieve the highest possible Ra. However, here we show that for very high Ra integral quantities such as the Nusselt number and volume averaged Reynolds number only converge to the large aspect ratio limit around Γ ≈4 , while horizontally averaged statistics such as standard deviation and kurtosis converge around Γ ≈8 , the integral scale converges around Γ ≈32 , and the peak position of the temperature variance and turbulent kinetic energy spectra only converge around Γ ≈64 .

  18. Nonlinear analysis of collapse mechanism in superstructure vehicle

    Science.gov (United States)

    Nor, M. K. Mohd; Ho, C. S.; Ma'at, N.

    2017-04-01

    The EU directive 2001/85/EC is an official European text which describes the specifications for "single deck class II and III vehicles" required to be approved by the regulation UN/ECE no.66 (R66). To prevent the catastrophic consequences by occupant during an accident, the Malaysian government has reinforced the same regulation upon superstructure construction. This paper discusses collapse mechanism analysis of a superstructure vehicle using a Crash D nonlinear analysis computer program based on this regulation. The analysis starts by hand calculation to define the required energy absorption by the chosen structure. Simple calculations were then performed to define the weakest collapse mechanism after undesirable collapse modes are eliminated. There are few factors highlighted in this work to pass the regulation. Using the selected cross section, Crash D simulation showed a good result. Generally, the deformation is linearly correlates to the energy absorption for the structure with low stiffness. Failure of critical members such as vertical lower side wall must be avoided to sustain safety of the passenger compartment and prevent from severe and fatal injuries to the trapped occupant.

  19. Control superstructure of rigid polyelectrolytes in oppositely charged hydrogels via programmed internal stress

    Science.gov (United States)

    Takahashi, Riku; Wu, Zi Liang; Arifuzzaman, Md; Nonoyama, Takayuki; Nakajima, Tasuku; Kurokawa, Takayuki; Gong, Jian Ping

    2014-08-01

    Biomacromolecules usually form complex superstructures in natural biotissues, such as different alignments of collagen fibres in articular cartilages, for multifunctionalities. Inspired by nature, there are efforts towards developing multiscale ordered structures in hydrogels (recognized as one of the best candidates of soft biotissues). However, creating complex superstructures in gels are hardly realized because of the absence of effective approaches to control the localized molecular orientation. Here we introduce a method to create various superstructures of rigid polyanions in polycationic hydrogels. The control of localized orientation of rigid molecules, which are sensitive to the internal stress field of the gel, is achieved by tuning the swelling mismatch between masked and unmasked regions of the photolithographic patterned gel. Furthermore, we develop a double network structure to toughen the hydrogels with programmed superstructures, which deform reversibly under large strain. This work presents a promising pathway to develop superstructures in hydrogels and should shed light on designing biomimetic materials with intricate molecular alignments.

  20. Challenges associated with extending spent fuel storage until reprocessing or disposal

    International Nuclear Information System (INIS)

    Carlsen, Brett; Saegusa, Toshiari; Wasinger, Karl; Grahn, Per; Wolff, Dietmar; Waters, Michael; Bevilacqua, Arturo

    2014-01-01

    Existing spent fuel storage (SFS) practices are the result of the past presumptions that an end point, e.g. sufficient reprocessing and/or disposal capacity, would be available within the short term (approximately 50 years). Consequently, long term storage (between approximately 50 and 100 years) considerations have not been included in planning the back end of the nuclear fuel cycle. The present reality shows that no country has yet neither licensed nor built nor operated a deep geological repository for spent fuel (SF) and/or high level waste (HLW). Further, present and projected SF generation rates - more than 10 000 metric tons of heavy metal (MTHM) a year - far exceed the current capacity for disposal - 0 MTHM - or reprocessing - 4 800 MTHM a year - and will continue to do so for the rest of this decade. As a result, the SFS periods will extend. Moreover, as the SFM end point - reprocessing and/or disposal - is not presently defined with certainty in most countries, SFS periods will extend over periods within or beyond the long term in those countries. The IAEA has started in October 2010 a programmatic activity to consider challenges associated with extending SFS durations. After four consultants meetings and two technical meetings, a need has been identified for a SFS framework based on renewable storage periods - with as many renewals as may be needed - to ensure safe and secure SFS until sufficient reprocessing and/or disposal capacity is implemented. Over the course of the technical meetings, the consultants have worked with delegates of 36 Member States and 2 International Organizations to emphasize the importance of establishing programs that can provide sufficient confidence that age-related degradation will be recognized and addressed to effectively prevent unacceptable consequences. This paper considers a number of topics from the perspective of assuring safe and effective SFS as storage periods extend including: SFS concepts, packaging of SF

  1. Characterization of the 309 building fuel transfer pit and storage basin

    International Nuclear Information System (INIS)

    Hale, N.S.

    1998-01-01

    This document identifies radiological, chemical and physical conditions inside the Fuel Transfer Pit and Fuel Storage Basins. These spaces are located inside the Plutonium Recycle Test Reactor structure (309 Building.) The fuel handling and storage feature of the PRTR were primarily located in these spaces. The conditions were assessed as part of overall 309 Building transition

  2. Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2014-10-15

    In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al{sup 27}, C{sup 12}, B{sup 11}, B{sup 10} and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B{sup 10} content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B{sup 10} content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B{sup 10} content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B{sup 10} content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10{sup 10} order, however, usual neutron flux from spent fuel is 10{sup 8} order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B{sup 10} content is little decreased, so, initial neutron absorbing ability could be kept continuously.

  3. Vision and framework for technical and management support to facilitate foreign spent fuel storage and geologic disposal in Russia

    International Nuclear Information System (INIS)

    Halsey, W.G.; Jardine, L.J.; Smith, C.F.

    1999-01-01

    This 'Technical and Management Support' program would facilitate the transfer of spent fuel from commercial power plants in Taiwan to a storage and geologic repository site near Krasnoyarsk, Russia. This program resolves issues of disposition of Taiwan spent fuel (including US origin fuel) and provides revenue for Russia to develop an integrated spent fuel storage and radioactive waste management system including a geologic repository. LLNL has ongoing contracts and collaborations with all the principal parties and is uniquely positioned to facilitate the development of such a program. A three-phase approach over 20 years is proposed: namely, an initial feasibility investigation followed by an engineering development phase, and then implementation

  4. Intervention in independent spent fuel storage facility license application proceedings for storage on the power plant site

    International Nuclear Information System (INIS)

    Jordan, J.

    1992-01-01

    This presentation summarizes the intervention in the Nuclear Regulatory Commission (NRC) licensing process for currently operating Independent Spent fuel Storage Installation (ISFSI) projects at Carolina Power and Light's Company's H.B. Robinson, Duke Power Company's Oconee, and Virginia Power Company's Surry. In addition, intervention at dry storage facilities that are currently under development are also described. The utilities and reactors include Baltimore Gas and Electric Company's Calvert Cliffs, Public Service Company of Colorado's Fort St. Vrain plant, Northern States Power Company's Prairie Island, Wisconsin Electric Power Company's Point Beach, and Consumers Power Company's Palisades

  5. Abnormal Congenital Location of Stapes’ Superstructure: Clinical and Embryological Implications

    Directory of Open Access Journals (Sweden)

    Vânia Henriques

    2016-01-01

    Full Text Available Congenital middle ear malformations are rare. Most part of them are usually associated with other malformations, such as aural atresia, microtia, and dysmorphic craniofacial features. A clinical case of a 24-year-old male with a right-sided conductive hearing loss since his childhood, without craniofacial malformation, is presented. He was proposed for exploratory tympanotomy under the suspicious diagnosis of otosclerosis. The surgery revealed an abnormal location of stapes’ superstructure, which was attached to the promontory and had an isolated and mobile osseous footplate in the oval window. A stapes prosthesis was inserted and resulted in closure of the air-bone gap by 25 dB. A review of the literature was also performed using MEDLINE. Two theories diverge on the embryologic origin of the stapes. Our findings seem to be in favour of the theory that defines two different embryologic origins to the stapes.

  6. Investigation of superstructure damping identification for the HDR containment building

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Kot, C.A.; Srinivasan, M.G.

    1985-01-01

    A method for the estimation of first mode structural damping, developed by other investigators, was applied to shaker test data of the HDR containment building. Due to inadequate precision in the experimental phase measurements no valid results could be obtained. Based on modal analysis it was also noted that for systems such as the HDR building, contributions of higher modes are not negligible as was assumed in the original approach. Therefore, the procedure for the determination of superstructure damping using experimental data was extended to include the effects of higher modes. The extended method does not lead to any higher order nonlinear equations than the first mode approximation and was found to be as simple to apply as the original approach

  7. Multi-objective superstructure-free synthesis and optimization of thermal power plants

    International Nuclear Information System (INIS)

    Wang, Ligang; Lampe, Matthias; Voll, Philip; Yang, Yongping; Bardow, André

    2016-01-01

    The merits of superstructure-free synthesis are demonstrated for bi-objective design of thermal power plants. The design of thermal power plants is complex and thus best solved by optimization. Common optimization methods require specification of a superstructure which becomes a tedious and error-prone task for complex systems. Superstructure specification is avoided by the presented superstructure-free approach, which is shown to successfully solve the design task yielding a high-quality Pareto front of promising structural alternatives. The economic objective function avoids introducing infinite numbers of units (e.g., turbine, reheater and feedwater preheater) as favored by pure thermodynamic optimization. The number of feasible solutions found per number of mutation tries is still high even after many generations but declines after introducing highly-nonlinear cost functions leading to challenging MINLP problems. The identified Pareto-optimal solutions tend to employ more units than found in modern power plants indicating the need for cost functions to reflect current industrial practice. In summary, the multi-objective superstructure-free synthesis framework is a robust approach for very complex problems in the synthesis of thermal power plants. - Highlights: • A generalized multi-objective superstructure-free synthesis framework for thermal power plants is presented. • The superstructure-free synthesis framework is comprehensively evaluated by complex bi-objective synthesis problems. • The proposed framework is effective to explore the structural design space even for complex problems.

  8. Criticality safety assessment on the RSG-GAS spent fuel storage for anticipating the next core conversion program

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Kuntoro, Iman; Zuhair; Liem, Peng Hong

    2003-01-01

    Criticality assessment on the spent fuel storage racks of the RSG-GAS multipurpose reactor has been conducted to support the undergoing core conversion program, in which higher uranium fuel densities of silicide (up to 4.8 gU.cm -3 ) and molybdenum (up to 8.3 gU.cm -3 ) fuel elements are adopted to enhance the reactor performance, core cycle length and reactor utilization. In the assessment, the k eff of the rack as a function of fuel density is calculated for fresh fuel elements which is a very conservative approach recommended by IAEA. Besides fuel densities, effects of water densities due to pool water temperature variation, and the fuel elements' orientation on the k eff are analyzed as well. The criticality calculations are all carried out by using MNCP4B2 Monte Carlo code with ENDF/B-VI library. For the library sensitivity, JENDL-3.3 library is also used and compared. The calculation results show the most reactive condition is for the case when the spent fuel racks are filled with fresh U-6Mo fuel element with meat density of 8.30 gU.cm -3 . For all fuel types, density and operating condition, the calculated k eff with 3 times standard deviations are confirmed less than the allowable value of 0.95. It can be concluded that the existing spent fuel storage racks can be safely used for storing the planned high density uranium fuels. (author)

  9. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    Thomas, Jay

    1999-01-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S.-origin from foreign research reactors to the United States. As of July 1999, over 18% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These 2400 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into L-Area in April 1997 and approximately 86 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment will be explained to show how the empty

  10. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Conatser, E.R.; Thomas, J.E. [Westinghouse Savannah River Company, Aiken, SC 29808 (United States)

    2000-07-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S. origin from foreign research reactors to the United States. As of December 1999, over 22% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These {approx}2650 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed the L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into the L-Area in April 1997 and approximately 105 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment

  11. Overview of the spent nuclear fuel storage facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    Conatser, E.R.; Thomas, J.E.

    2000-01-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy concerning Foreign Research Reactor Spent Nuclear Fuel initiated a 13 year campaign renewing a policy to support the return of spent nuclear fuel containing uranium of U.S. origin from foreign research reactors to the United States. As of December 1999, over 22% of the approximately 13,000 spent nuclear fuel assemblies from participating countries have been returned to the Savannah River Site (SRS). These ∼2650 assemblies are currently stored in two dedicated SRS wet storage facilities. One is the Receiving Basin for Off-site Fuels (RBOF) and the other as L-Basin. RBOF, built in the early 60's to support the 'Atoms for Peace' program, has been receiving off-site fuel for over 35 years. RBOF has received approximately 1950 casks since startup and has the capability of handling all of the casks currently used in the FRR program. However, RBOF is 90% filled to capacity and is not capable of storing all of the fuel to be received in the program. L-Basin was originally used as temporary storage for materials irradiated in SRS's L-Reactor. New storage racks and other modifications were completed in 1996 that improved water quality and allowed the L-Basin to receive, handle and store spent nuclear fuel assemblies and components from off-site. The first foreign cask was received into the L-Area in April 1997 and approximately 105 foreign and domestic casks have been received since that time. This paper provides an overview of activities related to fuel receipt and storage in both the Receiving Basin for Off-site Fuels (RBOF) and L-Basin facilities. It will illustrate each step of the fuel receipt program from arrival of casks at SRS through cask unloading and decontamination. It will follow the fuel handling process, from fuel unloading, through the cropping and bundling stages, and final placement in the wet storage rack. Decontamination methods and equipment will be explained to show

  12. Linking experiment and theory for three-dimensional networked binary metal nanoparticle–triblock terpolymer superstructures

    KAUST Repository

    Li, Zihui; Hur, Kahyun; Sai, Hiroaki; Higuchi, Takeshi; Takahara, Atsushi; Jinnai, Hiroshi; Gruner, Sol M.; Wiesner, Ulrich

    2014-01-01

    the intimate coupling of synthesis, in-depth electron tomographic characterization and theory enables exquisite control of superstructure in highly ordered porous three-dimensional continuous networks from single and binary mixtures of metal nanoparticles

  13. Large-scale laboratory observations of wave forces on a highway bridge superstructure.

    Science.gov (United States)

    2011-10-01

    The experimental setup and data are presented for a laboratory experiment conducted to examine realistic wave forcing on a highway bridge : superstructure. The experiments measure wave conditions along with the resulting forces, pressures, and struct...

  14. Effects of radiation and environmental factors on the durability of materials in spent fuel storage and disposal

    International Nuclear Information System (INIS)

    2002-12-01

    This is the second report that addresses results from the Coordinated Research Project (CRP) on Irradiation Enhanced Degradation of Materials in Spent Fuel Storage Facilities. This second report addresses results of topical studies that are relevant to issues important to materials behaviour in wet storage technology, but also involves topics on materials behaviour in dry storage and repository environments, including effects of radiation. The material is in seven separate papers contributed by the participants in the CRP and contains details of research studies started within the framework of the CRP and in several cases completed well after the CRP was finished. The seven contributions fall into three broad subject areas: Effects of temperature and radiation on aqueous and moist air corrosion of stainless steels; Studies of materials behaviour in wet and dry storage; Effects of gamma radiation on the durability of candidate canister materials for repository applications: carbon steel, titanium, and copper. Each of the papers has been indexed separately

  15. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    International Nuclear Information System (INIS)

    Hozer, Zoltan; Szabo, Emese; Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor; Vajda, Nora

    2009-01-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  16. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Zoltan, E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Szabo, Emese [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor [Nuclear Power Plant Paks, H-7031 Paks, P.O. Box 71 (Hungary); Vajda, Nora [Institute of Nuclear Techniques, Budapest University of Technology and Economics, H-1521 Budapest, Muegyetem rakpart 9 (Hungary)

    2009-07-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  17. Geological evaluation of spent fuel storage and low-intermediate level radwaste disposal in the site of NPP candidate

    International Nuclear Information System (INIS)

    Sucipta; Yatim, S.; Martono, H.; Pudyo, A.

    1997-01-01

    Based on the consideration of techno-economy and environmental safety, the radioactive waste treatment installation (RWI), interim storage of spen fuel (ISSF) and low-intermediate level disposal shall be sited in the surrounding of NPP area. The land suitability of NPP's site candidate at Muria Peninsula as spent fuel storage and low-intermediate level radwaste disposal need to be studied. Site selection was conducted by overlay method and scoring method, and based on safety criteria which include geological and environmental aspects. Land evaluation by overlay method has given result a potential site which have highest suitable land at surrounding of borehole L-15 about 17.5 hectares. Land evaluation by scoring method has given result two land suitability classes, i.e. moderate suitability class (includes 14 borehole) and high suitability class, include borehole L-2, L-14 and L-15 (author)

  18. Assessment of furnaces including fuel storage facilities according to the 12th Federal Emission Control Ordinance (BImSchV)

    International Nuclear Information System (INIS)

    Hensler, G.; Ott, H.; Wunderlich, O.; Mair, K.

    1990-01-01

    Existing quantities of substances pursuant to Annex II of the 12th Federal Emission Control Ordinance in furnaces or in fuel storage facilities do not present a general hazard for fireplaces fired with coal, wood, heavy and light fuel oil within the meaning of the Accident Ordinance. In case of a fire in a storage facility for black coal, brown coal, untreated wood, light and heavy fuel oil, a general hazard on account of the release of developed substances is obviously excluded. Dispersion calculations pursuant to VDI 3783 have shown that concentrations of beryllium, arsenic, nickel, cobalt and mercury compounds in the vicinity of the fire source are so small that a general hazard can be excluded. (orig./DG) [de

  19. Site-specific issues related to structural/seismic design of an underground independent spent fuel storage installation (ISFSI)

    International Nuclear Information System (INIS)

    Tripathi, B.P.

    2005-01-01

    Utilities owning and operating commercial nuclear power plants (NPP) in USA may choose to build an underground Independent Spent Fuel Storage Installation (ISFSI) to store the spent nuclear fuels. The regulatory requirements and other guidance are based on 10 CFR Part 72, Regulatory Guide RG 3.73, Standard Review Plans NUREG-1536 and NUREG-1567, and Interim staff Guidance (ISG) documents as applicable. Structures, Systems, and Components (SSCs) classified as important to safety are designed to withstand the effects of site-specific environmental conditions and natural phenomena such as earthquake, tornado, flood, etc. An underground ISFSI for storage of spent nuclear fuel, presents some unique analysis and design challenges. This paper will briefly address some of these challenges and discuss site-specific loads, including seismic for the ISFSI design. (authors)

  20. Status and operational experience report of spent fuel storage facility in Kozloduy NPP for the period 1990 - 1994

    Energy Technology Data Exchange (ETDEWEB)

    Kalimanov, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1994-12-31

    Spent Fuel Storage Facility (SFSF) of Kozloduy NPP is designed for a long-term storage of 4920 spent fuel assemblies which are generated by all units for ten year operational period. The assemblies are stored in SFSF after 3 year storage in the reactor cooling pool. The SFSF operational safety is ensured by a number of strictly followed regulations related to: arrangement of the assemblies and conditions at which they are stored; transportation of the assemblies to the facility; residual heat removal; quality of the water used in the storage pool; water temperature and level control. Two independent groups of experts have carried out investigations to study the building safety. Their reports have been considered and accepted by the council of the Ministry of Environment which was the final step in licensing the SFSF.

  1. Results on Technical and Consultants Service Meetings on Lessons Learned from Operating Experience in Wet and Dry Spent Fuel Storage

    International Nuclear Information System (INIS)

    White, B.; Zou, X.

    2015-01-01

    Spent fuel storage has been and will continue to be a vital portion of the nuclear fuel cycle, regardless of whether a member state has an open or closed nuclear fuel cycle. After removal from the reactor core, spent fuel cools in the spent fuel pool, prior to placement in dry storage or offsite transport for disposal or reprocessing. Additionally, the inventory of spent fuel at many reactors worldwide has or will reach the storage capacity of the spent fuel pool; some facilities are alleviating their need for additional storage capacity by utilizing dry cask storage. While there are numerous differences between wet and dry storage; when done properly both are safe and secure. The nuclear community shares lessons learned worldwide to gain knowledge from one another’s good practices as well as events. Sharing these experiences should minimize the number of incidents worldwide and increase public confidence in the nuclear industry. Over the past 60 years, there have been numerous experiences storing spent fuel, in both wet and dry mediums, that when shared effectively would improve operations and minimize events. These lessons learned will also serve to inform countries, who are new entrants into the nuclear power community, on designs and operations to avoid and include as best practices. The International Atomic Energy Agency convened a technical and several consultants’ meetings to gather these experiences and produce a technical document (TECDOC) to share spent fuel storage lessons learned among member states. This paper will discuss the status of the TECDOC and briefly discuss some lessons learned contained therein. (author)

  2. Bilateral Congenital Absence of the Stapes Superstructure in Two Siblings

    Directory of Open Access Journals (Sweden)

    Jose Ignacio Undabeitia

    2014-01-01

    Full Text Available Middle ear ossicle malformations are an uncommon event. Among them, the congenital absence of the stapes is a very rare condition that is seldom described in the literature. We report the cases of two women, aged 19 and 22 , who presented with a long history of conductive deafness. An exploratory tympanotomy was performed and the absence of the stapes superstructure and an abnormal position of the facial nerve could be observed. A bone anchored hearing aid (BAHA was implanted in both patients with good results. It is believed that stapes agenesis is related to an abnormal development of the facial nerve, which by the 5th to 6th week of gestation would interpose between the otic capsule and the stapes blastema, preventing these structures from contacting. A long history of nonprogressive hearing loss from birth or early childhood is the key to reach a diagnosis. Several treatment options have been described. The authors opted for a hearing aid due to the high risk of facial nerve lesion, with good functional results.

  3. K Basins fuel encapsulation and storage hazard categorization

    International Nuclear Information System (INIS)

    Porten, D.R.

    1994-12-01

    This document establishes the initial hazard categorization for K-Basin fuel encapsulation and storage in the 100 K Area of the Hanford site. The Hazard Categorization for K-Basins addresses the potential for release of radioactive and non-radioactive hazardous material located in the K-Basins and their supporting facilities. The Hazard Categorization covers the hazards associated with normal K-Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. The criteria categorizes a facility based on total curies per radionuclide located in the facility. Tables 5-3 and 5-4 display the results in section 5.0. In accordance with DOE-STD-1027 and the analysis provided in section 5.0, the K East Basin fuel encapsulation and storage activity and the K West Basin storage are classified as a open-quotes Category 2close quotes Facility

  4. Structural safety evaluation of the K Basin railcar and truck applications

    International Nuclear Information System (INIS)

    Winkel, B.V.

    1995-08-01

    There are two rail spurs in the storage/transfer areas of both the K East and K West fuel storage basins. These rail spurs both end at the west edge of the basins. To avoid accidental entry of a railcar into a basin, administrative procedures and rail control hardware have been provided. Based upon a combination of historical documentation and existing adminstrative controls, a maximum credible impact accident was established. Using this design basis accident, the existing rail control hardware was evaluated for structural adequacy. The K Basin rail spurs are embedded in concrete, which permits truck/trailer entry into the same area. Safety issues for truck applications are also addressed

  5. K East basin sludge volume estimates for integrated water treatment system

    International Nuclear Information System (INIS)

    Pearce, K.L.

    1998-01-01

    This document provides estimates of the volume of sludge expected from Integrated Process Strategy (IPS) processing of the fuel elements and in the fuel storage canisters in K East Basin. The original estimates were based on visual observations of fuel element condition in the basin and laboratory measurements of canister sludge density. Revision 1 revised the volume estimates of sludge from processing of the fuel elements based on additional data from evaluations of material from the KE Basin fuel subsurface examinations. A nominal Working Estimate and an upper level Working Bound is developed for the canister sludge and the fuel wash sludge components in the KE Basin

  6. Engineering study: 105KE to 105KW Basin fuel and sludge transfer. Final report

    International Nuclear Information System (INIS)

    Gant, R.G.

    1994-01-01

    In the last five years, there have been three periods at the 105KE fuel storage basin (KE Basin) where the reported drawdown test rates were in excess of 25 gph. Drawdown rates in excess of this amount have been used during past operations as the primary indicators of leaks in the basin. The latest leak occurred in March, 1993. The reported water loss from the KE Basin was estimated at 25 gph. This engineering study was performed to identify and recommend the most feasible and practical method of transferring canisters of irradiated fuel and basin sludge from the KE Basin to the 105KW fuel storage basin (KW Basin). Six alternatives were identified during the performance of this study as possible methods for transferring the fuel and sludge from the KE Basin to the KW Basin. These methods were then assessed with regard to operations, safety, radiation exposure, packaging, environmental concerns, waste management, cost, and schedule; and the most feasible and practical methods of transfer were identified. The methods examined in detail in this study were based on shipment without cooling water except where noted: Transfer by rail using the previously used transfer system and water cooling; Transfer by rail using the previously used transfer system (without water cooling); Transfer by truck using the K Area fuel transfer cask (K Area cask); Transfer by truck using a DOE shipping cask; Transfer by truck using a commercial shipping cask; and Transfer by truck using a new fuel shipping cask

  7. The Effects on the Operating Condition of a Passenger Ship Retro-fitted with a Composite Superstructure

    DEFF Research Database (Denmark)

    Karatzas, Vasileios; Hjørnet, N. K.; Kristensen, Hans Otto Holmegaard

    2016-01-01

    As sustainability and climate change have come on the politi-cal agenda, the shipping industry will have to be operating energy efficient ships. An appealing step to achieve this goal is by designing superstructures made out of Fiber Reinforced Plastics (FRP) aiming at the reduction of the ship......’s lightship weight. The benefits of a light superstructure become most prominent in large passenger ships, as the superstructures constitute a significant percentage of the lightship. Additional-ly, depending on the size of the ship, the superstructure may tower several decks above the weather deck, affecting...

  8. A novel ethanol templating synthesis of ordered lamellar superstructured crystalline zirconia

    Energy Technology Data Exchange (ETDEWEB)

    Liu Chao, E-mail: liuchao_tj@yahoo.com; Wang Bin [Key Laboratory for New Type of Functional Materials in Hebei Province, Hebei University of Technology (China); Ji Xiujie, E-mail: jxjchem@yahoo.com [State Key Laboratory of Hollow Fiber Membrane Materials and Processes, Tianjin Polytechnic University (China); Zhao Shanshan; Wu Jie; Jia Jianlong; Ma Dongxia [Key Laboratory for New Type of Functional Materials in Hebei Province, Hebei University of Technology (China)

    2012-03-15

    Soft template technique has attracted great interest, because it is a facile, inexpensive and efficient synthesis strategy for ordered superstructural systems. Here, a novel ethanol template was used to synthesize the ordered lamellar superstructured crystalline zirconia (L{alpha}-ZrO{sub 2}) without post-treatments and surfactants. ZrOCl{sub 2} and NaOH were served as Zr source and precipitant, respectively. XRD analysis showed that L{alpha}-ZrO{sub 2} is crystalline. XPS spectra indicated the physical adsorption of ethanol molecules in L{alpha}-ZrO{sub 2}. TEM further observed and proved the 1.36-nm period of superstructure detected and calculated by SAXRD (1.35 nm), which is composed of 0.68-nm thick ZrO{sub 2} and pore alternatively. In contrast, the template-free ZrO{sub 2} (TF-ZrO{sub 2}) presents no superstructure and is poorly crystallized. As a soft template, ethanol presents the roles of (i) inducing the growth of zirconia layers, (ii) directing the self-assembly of ordered lamellar superstructure, and (iii) decreasing the crystallization temperature. The possible mechanism of ethanol serving as a soft template was proposed and discussed in thermodynamics.

  9. A novel ethanol templating synthesis of ordered lamellar superstructured crystalline zirconia

    International Nuclear Information System (INIS)

    Liu Chao; Wang Bin; Ji Xiujie; Zhao Shanshan; Wu Jie; Jia Jianlong; Ma Dongxia

    2012-01-01

    Soft template technique has attracted great interest, because it is a facile, inexpensive and efficient synthesis strategy for ordered superstructural systems. Here, a novel ethanol template was used to synthesize the ordered lamellar superstructured crystalline zirconia (Lα-ZrO 2 ) without post-treatments and surfactants. ZrOCl 2 and NaOH were served as Zr source and precipitant, respectively. XRD analysis showed that Lα-ZrO 2 is crystalline. XPS spectra indicated the physical adsorption of ethanol molecules in Lα-ZrO 2 . TEM further observed and proved the 1.36-nm period of superstructure detected and calculated by SAXRD (1.35 nm), which is composed of 0.68-nm thick ZrO 2 and pore alternatively. In contrast, the template-free ZrO 2 (TF-ZrO 2 ) presents no superstructure and is poorly crystallized. As a soft template, ethanol presents the roles of (i) inducing the growth of zirconia layers, (ii) directing the self-assembly of ordered lamellar superstructure, and (iii) decreasing the crystallization temperature. The possible mechanism of ethanol serving as a soft template was proposed and discussed in thermodynamics.

  10. Analysis of sludge from Hanford K East Basin canisters

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J. [ed.] [comp.] [DE and S Hanford, Inc., Richland, WA (United States); Welsh, T.L. [B and W Protec, Inc. (United States); Baker, R.B. [DE and S Hanford, Inc., Richland, WA (United States); Hoppe, E.W.; Schmidt, A.J.; Abrefah, J.; Tingey, J.M.; Bredt, P.R.; Golcar, G.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-09-12

    Sludge samples from the canisters in the Hanford K East Basin fuel storage pool have been retrieved and analyzed. Both chemical and physical properties have been determined. The results are to be used to determine the disposition of the bulk of the sludge and to assess the impact of residual sludge on dry storage of the associated intact metallic uranium fuel elements. This report is a summary and review of the data provided by various laboratories. Although raw chemistry data were originally reported on various bases (compositions for as-settled, centrifuged, or dry sludge) this report places all of the data on a common comparable basis. Data were evaluated for internal consistency and consistency with respect to the governing sample analysis plan. Conclusions applicable to sludge disposition and spent fuel storage are drawn where possible.

  11. Turbulent Superstructures in Rayleigh-Bénard convection at different Prandtl number

    Science.gov (United States)

    Schumacher, Jörg; Pandey, Ambrish; Ender, Martin; Westermann, Rüdiger; Scheel, Janet D.

    2017-11-01

    Large-scale patterns of the temperature and velocity field in horizontally extended cells can be considered as turbulent superstructures in Rayleigh-Bénard convection (RBC). These structures are obtained once the turbulent fluctuations are removed by a finite-time average. Their existence has been reported for example in Bailon-Cuba et al.. This large-scale order obeys a strong similarity with the well-studied patterns from the weakly nonlinear regime at lower Rayleigh number in RBC. In the present work we analyze the superstructures of RBC at different Prandtl number for Prandtl values between Pr = 0.005 for liquid sodium and 7 for water. The characteristic evolution time scales, the typical spatial extension of the rolls and the properties of the defects of the resulting superstructure patterns are analyzed. Data are obtained from well-resolved spectral element direct numerical simulations. The work is supported by the Priority Programme SPP 1881 of the Deutsche Forschungsgemeinschaft.

  12. Linking experiment and theory for three-dimensional networked binary metal nanoparticle–triblock terpolymer superstructures

    KAUST Repository

    Li, Zihui

    2014-02-21

    © 2014 Macmillan Publishers Limited. Controlling superstructure of binary nanoparticle mixtures in three dimensions from self-assembly opens enormous opportunities for the design of materials with unique properties. Here we report on how the intimate coupling of synthesis, in-depth electron tomographic characterization and theory enables exquisite control of superstructure in highly ordered porous three-dimensional continuous networks from single and binary mixtures of metal nanoparticles with a triblock terpolymer. Poly(isoprene-block-styrene-block-(N,N-dimethylamino)ethyl methacrylate) is synthesized and used as structure-directing agent for ligand-stabilized platinum and gold nanoparticles. Quantitative analysis provides insights into short-and long-range nanoparticle-nanoparticle correlations, and local and global contributions to structural chirality in the networks. Results provide synthesis criteria for next-generation mesoporous network superstructures from binary nanoparticle mixtures for potential applications in areas including catalysis.

  13. Practical experience for liquid radioactive waste treatment from spent fuel storage pool on RA reactor in Vinca Institute

    International Nuclear Information System (INIS)

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2002-01-01

    The present paper reports the results of the preliminary removal of sludge from the bottom of the spent fuel storage pool in the RA reactor, mechanical filtration of the pool water and sludge conditioning and storage. Yugoslavia is a country without a nuclear power plant (NPP) on its territory. The law which strictly forbids NPP construction is still valid, but, nevertheless we must handle and dispose radioactive waste. This is not only because of radwaste originating from the use of radioactive materials in medicine and industry, but also because of the waste generated by research in the Nuclear Sciences Institute Vinca. In the last forty years, in the Vinca Institute, as a result of two research reactors being operational, named RA and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, radioactive waste materials are stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed in concrete containers, for further manipulation and disposal.(author)

  14. Maine Yankee: Making the Transition from an Operating Plant to an Independent Spent Fuel Storage Installation (ISFSI)

    International Nuclear Information System (INIS)

    Norton, W.; McGough, M. S.

    2002-01-01

    The purpose of this paper is to describe the challenges faced by Maine Yankee Atomic Power Company in making the transition from an operating nuclear power plant to an Independent Spent Fuel Storage Installation (ISFSI). Maine Yankee (MY) is a 900-megawatt Combustion Engineering pressurized water reactor whose architect engineer was Stone and Webster. Maine Yankee was put into commercial operation on December 28, 1972. It is located on an 820-acre site, on the shores of the Back River in Wiscasset, Maine about 40 miles northeast of Portland, Maine. During its operating life, it generated about 1.2 billion kilowatts of power, providing 25% of Maine's electric power needs and serving additional customers in New England. Maine Yankee's lifetime capacity factor was about 67% and it employed more than 450 people. The decision was made to shutdown Maine Yankee in August of 1997, based on economic reasons. Once this decision was made planning began on how to accomplish safe and cost effective decommissioning of the plant by 2004 while being responsive to the community and employees

  15. From passive storage to daily waste retrieval; changing the working culture at Sellafield's pile fuel storage pond

    Energy Technology Data Exchange (ETDEWEB)

    Carlisle, Derek [Sellafield Ltd, Seascale, Cumbria (United Kingdom)

    2013-07-01

    The Pile Fuel Storage Pond (PFSP) was built in 1948/50 to treat materials from the Windscale Piles. Multiple operational regimes over the intervening 60 years have resulted in a complex inventory of spent nuclear fuels, solid and liquid intermediate level wastes. A coordinated programme of work, designed to retrieve and safely dispose of the pond contents, has been implemented to enable the decommissioning of the facility. The long period of passive storage operations which preceded the implementation of the programme meant that the operator was faced with a dual challenge of providing new technical capability and changing a working culture that was inappropriate for the dynamic environment required to successfully deliver the programme. It was recognised that the nature of the programme meant that implementing a standard manufacturing approach to operations would not be appropriate. In order to create a dynamic retrievals focussed working culture, the operator has vigorously embraced change programmes aimed at improving a number of working practices including encouraging innovation, managing integrated but flexible production schedules, and encouraging work-face problem solving. The combined impact of beginning to resolve the technical challenges and focussing on the delivery culture has resulted in the facility making a step change towards becoming fully retrievals operations focussed. (authors)

  16. From passive storage to daily waste retrieval; changing the working culture at Sellafield's pile fuel storage pond

    International Nuclear Information System (INIS)

    Carlisle, Derek

    2013-01-01

    The Pile Fuel Storage Pond (PFSP) was built in 1948/50 to treat materials from the Windscale Piles. Multiple operational regimes over the intervening 60 years have resulted in a complex inventory of spent nuclear fuels, solid and liquid intermediate level wastes. A coordinated programme of work, designed to retrieve and safely dispose of the pond contents, has been implemented to enable the decommissioning of the facility. The long period of passive storage operations which preceded the implementation of the programme meant that the operator was faced with a dual challenge of providing new technical capability and changing a working culture that was inappropriate for the dynamic environment required to successfully deliver the programme. It was recognised that the nature of the programme meant that implementing a standard manufacturing approach to operations would not be appropriate. In order to create a dynamic retrievals focussed working culture, the operator has vigorously embraced change programmes aimed at improving a number of working practices including encouraging innovation, managing integrated but flexible production schedules, and encouraging work-face problem solving. The combined impact of beginning to resolve the technical challenges and focussing on the delivery culture has resulted in the facility making a step change towards becoming fully retrievals operations focussed. (authors)

  17. An integrated approach to risk-based remediation of a former bulk fuel storage facility adjacent a marine environment

    Energy Technology Data Exchange (ETDEWEB)

    Kemp, L.; Hers, I. [Golder Associates Ltd., Vancouver, BC (Canada)

    2006-07-01

    An integrated approach to risk-based remediation of a former bulk fuel storage facility adjacent to a marine environment was discussed. The presentation provided an introduction and illustration to the site location and history, located close to Skagway, Alaska and northwestern British Columbia. The site investigation and conceptual model were also presented. The remedial approach was also described with reference to a risk-based action approach, remedial objectives, soil vapour extraction (SVE)-bioventing, and air sparging-biosparging. The objectives were to minimize potential exposure to aquatic receptors by minimizing non-aqueous phase liquids (NAPL) mobility and dissolved transport of petroleum hydrocarbons. Groundwater modeling to assess the attenuation rate and to determine remedial targets was also discussed. Model validation and results of groundwater modeling as well as remediation system details and performance were then provided. It was determined that significant attenuation is occurring and that effective mass removal and concentrations have been decreasing over time. It was demonstrated that risk-based remedial goals and hydrogeology can change with land use/development. tabs., figs.

  18. Development of advanced spent fuel management process / criticality safety analysis for integrated mockup and metallized spent fuel storage

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Shin, Hee Sung; Shin, Young Joon; Bae, Kang Mok

    1999-02-01

    Benchmark calculation for SCALE4.3 CSAS6 module and burnup credit criticality analysis performed by CSAS6 module are described in this report. Calculation biases by the SCALE4.3 CSAS6 module for PWR spent fuel, metallized spent fuel and aqueous nuclear materials have been determined on the basis of the benchmark to be 0.011, 0.023 and 0.010, respectively. The maximum allowable multiplication factor for an integrated mockup and metallized spent fuel storage is conservatively determined to be 0.927. With the aid of this code system, K eff values as a function of metallization ratio for the integrated mockup have been calculated. The maximum values of K eff for normal and hypothetical accident conditions are 0.346 and 0.598, respectively, much less than the maximum allowable multiplication factor of 0.927. Besides, burnup credit criticality analysis has been performed for infinite arrays of square and hexagonal canisters containing metallized spent fuel rods with different canister wall thickness, canister surface-to-surface distance and water content. It is revealed that the effective multiplication factor for canister arrays as mentioned above is well below the subcritical limit regardless of external conditions when its wall thickness is over 9 mm. (Author). 37 refs., 27 tabs., 64 figs

  19. Cryo-transmission electron microscopy of a superstructure of fluid dioleoylphosphatidylcholine (DOPC) membranes

    DEFF Research Database (Denmark)

    Klösgen, B; Helfrich, W

    1997-01-01

    a porous technical membrane. Sampling and cryofixation took place at various times within 3 weeks after the preparation. From the micrographs we infer that the small fraction of vesicles enclosing one another develop passages (connections) between the bilayers. In contrast, the superstructure is basically...... a feature of disconnected membranes. Among its modifications are isolated membrane bends or folds and a grainy membrane texture with a minimal grain spacing of 4-6 nm. In the extruded dispersions the passages and the superstructure seem to be formed mostly within the first day. The fraction of smooth...

  20. Antiphase boundaries, inversion, and ferroelastic domains in the striped-type superstructure of γ-brass Cu-Al alloys

    Science.gov (United States)

    Koyama, Y.; Hatano, M.; Tanimura, M.

    1996-05-01

    Features of the striped-type superstructure in γ-brass Cu-Al alloys have been investigated mainly by transmission electron microscopy in order to understand the role of a rhombohedral distortion in its stability. From dark field images taken from alloys exhibiting the superstructure, it was found that there exist two types of ferroelastic rhombohedral domains, which are in a twin relation. In other words, the superstructure is characterized by a periodic array of the ferroelastic domains as well as inversion antiphase boundaries. Because charge density waves should be responsible for the formation of the superstructure, as suggested in our previous paper [Phys. Rev. B 40, 5378 (1989)], the rhombohedral distortion must play a crucial role in their appearance. Presumably the distortion enlarges parallel portions of the Fermi surface. On the basis of the present experimental data, the interplay between the striped-type superstructure and the rhombohedral γ-brass structure is also discussed.

  1. Safety and engineering aspects of spent fuel storage. Proceedings of an international symposium held in Vienna, 10-14 October 1994

    International Nuclear Information System (INIS)

    1995-01-01

    Spent fuel management is one of the most vital, and common, problems for countries with nuclear reactors. In the closed nuclear fuel cycle, further storage capacity may be required to match the arisings of spent fuel with the available capacity of reprocessing plants. With respect to the once-through cycle, interim storage of spent fuel is required until a final repository has been constructed and is in service. Though present spent fuel storage technologies provide adequate protection to the population and the environment, there is keen interest in seeing whether further improvements can be achieved in the area of spent fuel storage. This Symposium was a co-operative effort between the Nuclear Energy Agency of the OECD and the IAEA. Such meetings have been organized once every four years since 1987. The purpose is to: exchange information on the state of the art and on prospects for spent fuel storage; discuss the worldwide situation and the major factors influencing national policies in this field; and identify the most important direction that national efforts and international co-operation should take in this area. Over 140 participants from 39 countries and 4 international organizations attended the Symposium. Thirty-five full papers and 13 posters were presented and are included in these Proceedings. Refs, figs and tabs

  2. Experimental Study of Properties of Pervious Concrete used for Bridge Superstructure

    DEFF Research Database (Denmark)

    Lund, Mia Schou Møller; Hansen, Kurt Kielsgaard; Hertz, Kristian Dahl

    2014-01-01

    around freezing point are exposed to a harsh freeze-thaw impact, since bridges are not only cooled from the topside, as a normal road bed, but also from underneath. Hence, the demands to the moisture properties of the superstructure are strict in order to ensure the necessary durability. Pearl...

  3. Erection of the Rokko island bridge superstructure. Rokko island kyojobu koji no seko

    Energy Technology Data Exchange (ETDEWEB)

    Tanino, A; Sakai, Y; Morikawa, Y; Sakoda, H; Egami, T; Maeda, Y; Shikata, H; Miyazaki, K [Kawasaki Heavy Industries, Ltd., Tokyo (Japan)

    1993-10-20

    Erection of the superstructure of the Rokko Island Bridge (220 m long) was outlined which connects Rokko Island with Kobe City. The weight of the superstructure was nearly 7,300 tons, the highest single weight of any unit bridge structures ever installed in Japan. Because various ships cruised at the erection site every day, the superstructure was assembled and painted in a factory, and after shipped to its destination on a large barge, it was hoisted onto the previously constructed piers by three giant floating cranes (FC). Since it was most essential to match hook loads among the FCs, the hook load control system was developed in which hook load data of each FC were sent to a control center every 4-5 seconds by wireless transmission, and the data as well as their required control limits were graphically displayed to support a situation recognition, resulting in rapid and safe erection. The erection work started from that of piers in Aug. 1992, and that of the superstructure started in Nov. 1992 and ended in Feb. 1993. 13 figs., 4 tabs.

  4. Self-assembly of polyhedral metal–organic framework particles into three-dimensional ordered superstructures

    NARCIS (Netherlands)

    Avci, Civan; Imaz, Inhar; Carné-Sánchez, Arnau; Pariente, Jose Angel; Tasios, Nikos; Pérez-Carvajal, Javier; Alonso, Maria Isabel; Blanco, Alvaro; Dijkstra, M.; López, Cefe; Maspoch, Daniel

    Self-assembly of particles into long-range, three-dimensional, ordered superstructures is crucial for the design of a variety of materials, including plasmonic sensing materials, energy or gas storage systems, catalysts and photonic crystals. Here, we have combined experimental and simulation data

  5. Role of Steric Hindrance in the Crystal Packing of Z′ = 4 Superstructure of Trimethyltin Hydroxide

    KAUST Repository

    Dey, S.

    2018-01-22

    The roomerature crystal structure of trimethyltin hydroxide, (CH)SnOH, has been described by Anderson et al. [Cryst. Growth Des. 2011, 11, 820-826] as a 2a × 2b × 8c, 32-fold superstructure. We report a a × b × 8c, eight-fold superstructure with orthorhombic P2cn symmetry and Z′ = 4. Structured diffuse scattering observed at the positions of presumed superlattice reflections along a∗ and b∗ might have appeared as Bragg reflections in the experiment by Anderson et al. Alternatively, Anderson et al. and the present work might have studied different polymorphs of (CH)SnOH. Crystalline (CH)SnOH constitutes polymeric chains arranged parallel to c. In the eight-fold superstructure at 220 K, the polymeric chains possess a distorted zigzag arrangement of linked linear O-Sn-O units with bent angle at oxygen of ∼139.2°. This structure is essentially different from the 8-helical arrangement in the published 32-fold superstructure model. The origin of the distorted zigzag structure is explained by steric hindrance between hydrogen atoms of adjacent hydroxy groups and (CH)Sn groups. Frustration in the packing of the chains is determined by steric hindrance between methyl groups of neighboring chains, which prevents the formation of interchain C-H···O hydrogen bonds.

  6. Globalisation in Africa: Reflecting on Peter Jarvis's Superstructure and Substructure Model

    Science.gov (United States)

    Preece, Julia

    2017-01-01

    This paper reflects on Peter Jarvis' book "Globalisation, lifelong learning and the learning society," volume 2--in which he describes human learning within a global context and factors contributing to globalisation. He describes the relationship of power between countries manifested as the superstructure and sub structure. The paper…

  7. A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hee Dong

    2012-02-15

    In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

  8. A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

    International Nuclear Information System (INIS)

    Sohn, Hee Dong

    2012-02-01

    In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

  9. Unknown Aspects of Self-Assembly of PbS Microscale Superstructures

    Science.gov (United States)

    Querejeta-Fernández, Ana; Hernández-Garrido, Juan C.; Yang, Hengxi; Zhou, Yunlong; Varela, Aurea; Parras, Marina; Calvino-Gámez, José J.; González-Calbet, Jose M.; Green, Peter F.; Kotov, Nicholas A.

    2012-01-01

    A lot of interesting and sophisticated examples of nanoparticle (NP) self-assembly (SA) are known. From both fundamental and technological standpoints this field requires advancements in three principle directions: a) understanding the mechanism and driving forces of three-dimensional (3D) SA with both nano- and micro-levels of organization; b) understanding of disassembly/deconstruction processes; and c) finding synthetic methods of assembly into continuous superstructures without insulating barriers. From this perspective, we investigated the formation of well-known star-like PbS superstructures and found a number of previously unknown or overlooked aspects that can advance the knowledge of NP self-assembly in these three directions. The primary one is that the formation of large seemingly monocrystalline PbS superstructures with multiple levels of octahedral symmetry can be explained only by SA of small octahedral NPs. We found five distinct periods in the formation PbS hyperbranched stars: 1) nucleation of early PbS NPs with an average diameter of 31 nm; 2) assembly into 100–500 nm octahedral mesocrystals; 3) assembly into 1000–2500 nm hyperbranched stars; 4) assembly and ionic recrystallization into six-arm rods accompanied by disappearance of fine nanoscale structure; 5) deconstruction into rods and cubooctahedral NPs. The switches in assembly patterns between the periods occur due to variable dominance of pattern–determining forces that include vander Waals and electrostatic (charge-charge, dipole-dipole, and polarization) interactions. The superstructure deconstruction is triggered by chemical changes in the deep eutectic solvent (DES) used as the media. PbS superstructures can be excellent models for fundamental studies of nanoscale organization and SA manufacturing of (opto)electronics and energy harvesting devices which require organization of PbS components at multiple scales. PMID:22515512

  10. Motions in Nearby Galaxy Cluster Reveal Presence of Hidden Superstructure

    Science.gov (United States)

    2004-09-01

    A nearby galaxy cluster is facing an intergalactic headwind as it is pulled by an underlying superstructure of dark matter, according to new evidence from NASA's Chandra X-ray Observatory. Astronomers think that most of the matter in the universe is concentrated in long large filaments of dark matter and that galaxy clusters are formed where these filaments intersect. A Chandra survey of the Fornax galaxy cluster revealed a vast, swept-back cloud of hot gas near the center of the cluster. This geometry indicates that the hot gas cloud, which is several hundred thousand light years in length, is moving rapidly through a larger, less dense cloud of gas. The motion of the core gas cloud, together with optical observations of a group of galaxies racing inward on a collision course with it, suggests that an unseen, large structure is collapsing and drawing everything toward a common center of gravity. X-ray Image of Fornax with labels X-ray Image of Fornax with labels "At a relatively nearby distance of about 60 million light years, the Fornax cluster represents a crucial laboratory for studying the interplay of galaxies, hot gas and dark matter as the cluster evolves." said Caleb Scharf of Columbia University in New York, NY, lead author of a paper describing the Chandra survey that was presented at an American Astronomical Society meeting in New Orleans, LA. "What we are seeing could be associated directly with the intergalactic gas surrounding a very large scale structure that stretches over millions of light years." The infalling galaxy group, whose motion was detected by Michael Drinkwater of the University of Melbourne in Australia, and colleagues, is about 3 million light years from the cluster core, so a collision with the core will not occur for a few billion years. Insight as to how this collision will look is provided by the elliptical galaxy NGC 1404 that is plunging into the core of the cluster for the first time. As discussed by Scharf and another group

  11. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  12. Spent fuel storage

    International Nuclear Information System (INIS)

    Huppert

    1976-01-01

    To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR) [de

  13. Fuel storage rack

    International Nuclear Information System (INIS)

    Mollon, L.

    1977-01-01

    Disclosed is a storage rack for spent nuclear fuel elements comprising a multiplicity of elongated hollow containers of uniform cross-section, preferably square,some of said containers having laterally extending continuous flanges extending between adjacent containers and defining continuous elongated chambers therebetween for the reception of neutron absorbing panels. 18 claims, 7 figures

  14. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Murakami, Naoshi.

    1996-01-01

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

  15. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Morikawa, Matsuo; Uchiyama, Yuichi.

    1983-01-01

    Purpose: To improve the safety and facilitate the design by limiting the relative displacement in a storage rack. Constitution: The outer wall of a storage rack disposed in water within a fuel pool, the pool wall opposing to the storage rack and the structure between the opposing storages racks are made as a space for confining the pool water or a structure formed with a slight gap, for example, a combination of a recessed structure and a protruded structure. In such a constitution, a space for confirming the pool water is established and the pool water thus confined forms a flow resistance when the storage rack vibrates upon earthquakes, serves as a damper and significantly reduces the responsivity. Furthermore, the relative displacement in the storage rack is limited to inhibit excess earthquake forces to exert on setting bolts and rack clamping bolts of the storage rack. (Sekiya, K.)

  16. Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Badwan, Faris M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott F [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-06

    Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the

  17. High-resolution electron microscopy on incommensurate long-period superstructures of hexagonal-close-packed Cu-Sb alloy

    International Nuclear Information System (INIS)

    Onozuka, T.; Kakehashi, S.; Takahashi, T.; Hirabayashi, M.

    1989-01-01

    Hexagonal incommensurate long-period superstructures of the Cu-Sb alloys containing 18-20 at.% Sb have been investigated by means of superstructure imaging using a high-resolution electron microscope. Honeycomb-type distributions of hexagonal domains consisting of the commensurate superstructure of type 7a 0 -2H are observed. The incommensurabilities of superstructure can be interpreted well with a hexagonal model composed of the 7a 0 -2H domains surrounded by domain walls which contain higher Sb content than the domain interior. The observed image contrast is reproduced well with multislice computer simulations based on the structure models proposed for the 7a 0 -2H domain and the domain wall. (orig.)

  18. High-resolution electron microscopy on incommensurate long-period superstructures of hexagonal-close-packed Cu-Sb alloy

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, T.; Kakehashi, S.; Takahashi, T.; Hirabayashi, M. (Tohoku Univ., Sendai (Japan). Inst. for Materials Research)

    1989-06-01

    Hexagonal incommensurate long-period superstructures of the Cu-Sb alloys containing 18-20 at.% Sb have been investigated by means of superstructure imaging using a high-resolution electron microscope. Honeycomb-type distributions of hexagonal domains consisting of the commensurate superstructure of type 7a{sub 0}-2H are observed. The incommensurabilities of superstructure can be interpreted well with a hexagonal model composed of the 7a{sub 0}-2H domains surrounded by domain walls which contain higher Sb content than the domain interior. The observed image contrast is reproduced well with multislice computer simulations based on the structure models proposed for the 7a{sub 0}-2H domain and the domain wall. (orig.).

  19. Status of nuclear fuel reprocessing, spent fuel storage, and high-level waste disposal. Nuclear Fuel Cycle Committee, California Energy Resources Conservation and Development Commission. Draft report

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    An analysis of the current status of technologies and issues in the major portions of the back-end of the nuclear fuel cycle is presented. The discussion on nuclear fuel reprocessing covers the reprocessing requirement, reprocessing technology assessment, technology for operation of reprocessing plants, and approval of reprocessing plants. The chapter devoted to spent fuel storage covers the spent fuel storge problem, the legislative response, options for maintaining full core discharge capacity, prospective availability of alterntive storage options, and the outlook for California. The existence of a demonstrated, developed high-level waste disposal technology is reviewed. Recommendations for Federal programs on high-level waste disposal are made

  20. DEPENDENCIES TO DETERMINE THE MEASURE OF DAMAGE AND CALCULATION OF RESIDUAL LIFE OF REINFORCED CONCRETE SUPERSTRUCTURE, EXPOSED TO SALT CORROSION

    OpenAIRE

    SAATOVA NODIRA ZIYAYEVNA

    2016-01-01

    In this paper we consider the current method of determining the measure of damage of concrete and reinforcement. The proposed dependence measures of damage, convenient for use in predicting the life of structures superstructures.The practical method of calculation determination of residual resource of the exploited superstructures developed. The main source of data for calculating the residual life are the parameters defined by the technical diagnosis.

  1. Optimal processing pathway for the production of biodiesel from microalgal biomass: A superstructure based approach

    DEFF Research Database (Denmark)

    Rizwan, Muhammad; Lee, Jay H.; Gani, Rafiqul

    2013-01-01

    In this study, we propose a mixed integer nonlinear programming (MINLP) model for superstructure based optimization of biodiesel production from microalgal biomass. The proposed superstructure includes a number of major processing steps for the production of biodiesel from microalgal biomass......, such as the harvesting of microalgal biomass, pretreatments including drying and cell disruption of harvested biomass, lipid extraction, transesterification, and post-transesterfication purification. The proposed model is used to find the optimal processing pathway among the large number of potential pathways that exist...... for the production of biodiesel from microalgae. The proposed methodology is tested by implementing on a specific case with different choices of objective functions. The MINLP model is implemented and solved in GAMS using a database built in Excel. The results from the optimization are analyzed...

  2. Basis for category B designation for K basins

    International Nuclear Information System (INIS)

    Jensen, M.A.

    1994-01-01

    This Supporting Document analyzes the various fissile material configurations in the 105-K East and K West fuel storage basins to determine the proper firefighting category. Firefighting categories are assigned to fissionable material facilities to provide guidance to firefighters in the allowable uses of water and other extinguishing materials to prevent inadvertent rearrangement of fissile materials or addition of neutron moderators which could lead to a criticality. This document concludes the appropriate category is B, which does not impose any restrictions on the use of water for firefighting purposes

  3. Design criteria document, Fire Protection Task, K Basin Essential Systems Recovery, Project W-405

    International Nuclear Information System (INIS)

    Johnson, B.H.

    1994-01-01

    The K Basin were constructed in the early 1950's with a 20 year design life. The K Basins are currently in their third design life and are serving as a near term storage facility for irradiated N Reactor fuel until an interim fuel storage solution can be implemented. In April 1994, Project W-405, K Basin Essential Systems Recovery, was established to address (among other things) the immediate fire protection needs of the 100K Area. A Fire Barrier Evaluation was performed for the wall between the active and inactive areas of the 105KE and 105KW buildings. This evaluation concludes that the wall is capable of being upgraded to provide an equivalent level of fire resistance as a qualified barrier having a fire resistance rating of 2 hours. The Fire Protection Task is one of four separate Tasks included within the scope of Project W405, K Basin Essential systems Recovery. The other three Tasks are the Water Distribution System Task, the Electrical System Task, and the Maintenance Shop/Support Facility Task. The purpose of Project W-405's Fire Protection Task is to correct Life Safety Code (NFPA 101) non-compliances and to provide fire protection features in Buildings 105KE, 105KW and 190KE that are essential for assuring the safe operation and storage of spent nuclear fuel at the 100K Area Facilities' Irradiated Fuel Storage Basins (K Basins)

  4. Enigmatic cranial superstructures among Chamorro ancestors from the Mariana Islands: gross anatomy and microanatomy.

    Science.gov (United States)

    Heathcote, Gary M; Bromage, Timothy G; Sava, Vincent J; Hanson, Douglas B; Anderson, Bruce E

    2014-06-01

    This study focuses on the gross anatomy, anatomic relations, microanatomy, and the meaning of three enigmatic, geographically patterned, and quasi-continuous superstructures of the posterior cranium. Collectively known as occipital superstructures (OSSs), these traits are the occipital torus tubercle (TOT), retromastoid process (PR), and posterior supramastoid tubercle (TSP). When present, TOT, PR, and TSP develop at posterior cranial attachment sites of the upper trapezius, superior oblique, and sternocleidomastoid muscles, respectively. Marked expression and co-occurrence of these OSSs are virtually circumscribed within Oceania and reach highest recorded frequencies in protohistoric Chamorros (CHamoru) of the Mariana Islands. Prior to undertaking scanning electron microscopy (SEM) work, our working multifactorial model for OSS development was that early-onset, long-term, and chronic activity-related microtrauma at enthesis sites led to exuberant reactive or reparative responses in a substantial minority of genetically predisposed (and mostly male) individuals. SEM imaging, however, reveals topographic patterning that questions, but does not negate, activity induction of these superstructures. Although OSSs appear macroscopically as relatively large and discrete phenomena, SEM findings reveal a unique, widespread, and seemingly systemic distribution of structures over the occipital surface that have the appearance of OSS microforms. Nevertheless, apparent genetic underpinnings, anatomic relationships with muscle entheses, and positive correlation of OSS development with humeral robusticity continue to suggest that these superstructures have potential to at once bear witness to Chamorro population history and inform osteobiographical constructions of chronic activity patterns in individuals bearing them. Further work is outlined that would illuminate the proximate and ultimate meanings of OSS. Copyright © 2014 Wiley Periodicals, Inc.

  5. Effect of modeling of super-structure on the behaviour of reactor building raft

    International Nuclear Information System (INIS)

    Mondal, A.; Singh, A.K.; Roy, Raghupati; Verma, U.S.P.; Warudkar, A.S.

    2003-01-01

    The behaviour of the reactor building raft was studied when the stiffness of the super-structural elements is included in the analysis as compared to the results of conventional analysis ignoring the stiffness of the super-structural elements. The effect of the stiffness of the super-structures on the loss of contact of the raft under seismic environment was also investigated. In order to study the effect of horizontal springs on the behaviour of the raft particularly near the stressing gallery under seismic environment, a separate study has been carried out considering a 3D model consisting of solid elements supported on both horizontal and vertical springs. The model was analysed for all the forces applied at the top of the raft and the analysis results were compared with those of shell model. The following conclusions are drawn: (i) Idealisation of the reactor building raft using shell elements is adequate for estimating the design forces/moments on the raft. The design forces/moments obtained from FE model consisting of solid elements closely matches with those obtained from FE model with shell elements. Idealisation of the RB raft using shell elements will also reduce the problem size and the related computational efforts. (ii) The stiffness of the super-structure has significant effect on the behaviour of the raft. Consideration of the stiffness of the super structure reduces the design forces/moments significantly and hence, modelling of the stiffness of the super structure is necessary for economical design. (iii) Modelling of horizontal stiffness of the raft in terms of horizontal springs at the interface of the raft and the rock does not have significant effect on the behaviour of the raft and as such, is not required to be considered in the FE model. However, it is necessary to ensure adequate factor of safety against the overall stability of the raft

  6. Automatic recognition of ship types from infrared images using superstructure moment invariants

    Science.gov (United States)

    Li, Heng; Wang, Xinyu

    2007-11-01

    Automatic object recognition is an active area of interest for military and commercial applications. In this paper, a system addressing autonomous recognition of ship types in infrared images is proposed. Firstly, an approach of segmentation based on detection of salient features of the target with subsequent shadow removing is proposed, as is the base of the subsequent object recognition. Considering the differences between the shapes of various ships mainly lie in their superstructures, we then use superstructure moment functions invariant to translation, rotation and scale differences in input patterns and develop a robust algorithm of obtaining ship superstructure. Subsequently a back-propagation neural network is used as a classifier in the recognition stage and projection images of simulated three-dimensional ship models are used as the training sets. Our recognition model was implemented and experimentally validated using both simulated three-dimensional ship model images and real images derived from video of an AN/AAS-44V Forward Looking Infrared(FLIR) sensor.

  7. Overcoming artificial spatial correlations in simulations of superstructure domain growth with parallel Monte Carlo algorithms

    International Nuclear Information System (INIS)

    Schleier, W.; Besold, G.; Heinz, K.

    1992-01-01

    The authors study the applicability of parallelized/vectorized Monte Carlo (MC) algorithms to the simulation of domain growth in two-dimensional lattice gas models undergoing an ordering process after a rapid quench below an order-disorder transition temperature. As examples they consider models with 2 x 1 and c(2 x 2) equilibrium superstructures on the square and rectangular lattices, respectively. They also study the case of phase separation ('1 x 1' islands) on the square lattice. A generalized parallel checkerboard algorithm for Kawasaki dynamics is shown to give rise to artificial spatial correlations in all three models. However, only if superstructure domains evolve do these correlations modify the kinetics by influencing the nucleation process and result in a reduced growth exponent compared to the value from the conventional heat bath algorithm with random single-site updates. In order to overcome these artificial modifications, two MC algorithms with a reduced degree of parallelism ('hybrid' and 'mask' algorithms, respectively) are presented and applied. As the results indicate, these algorithms are suitable for the simulation of superstructure domain growth on parallel/vector computers. 60 refs., 10 figs., 1 tab

  8. Seismic evaluation of K basin bridge cranes (HOI-320 ampersand HOI-418) and supporting structure

    International Nuclear Information System (INIS)

    Winkel, B.V.; Kanjilad, S.K.

    1996-03-01

    The Safety Class 1 100-K fuel storage basins are vulnerable to impact damage if a bridge crane were to fall during a seismic event. The pupose of this report is to address the adequacy of the K Basin bridge cranes to resist a seismic-induced fall. The approach used to demonstrate adequacy against falling, was to evaluate the crane structural components relative to requirements specified in ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes. Additionally, wheel lift-off and the adequacy of the crane supporting structure, are addressed. Seismic adequacy of the mechanical hoist equipment is not addressed in this report

  9. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  10. Morphology-controlled hydrothermal synthesis of MnCO3 hierarchical superstructures with Schiff base as stabilizer

    International Nuclear Information System (INIS)

    Hu, He; Xu, Jie-yan; Yang, Hong; Liang, Jie; Yang, Shiping; Wu, Huixia

    2011-01-01

    Graphical abstract: MnCO3 microcrystals with hierarchical superstructures were synthesized by using the CO2 in atmosphere as carbonate ions source and Schiff base as shape guiding-agent in water/ethanol system under hydrothermal condition. Highlights: → The most interesting in this work is the use of the greenhouse gases CO 2 in atmosphere as carbonate ions source to precipitate with Mn 2+ for producing MnCO 3 crystals. → This work is the first report related to the small organic molecule Schiff base as shape guiding-agent to produce different MnCO 3 hierarchical superstructures. → We are controllable synthesis of the MnCO 3 hierarchical superstructures such as chrysanthemum, straw-bundle, dumbbell and sphere-like microcrystals. → The as-prepared MnCO 3 could be used precursor to fabricate the Mn 2 O 3 hierarchical superstructures after thermal decomposition at high temperature. -- Abstract: MnCO 3 with hierarchical superstructures such as chrysanthemum, straw-bundle, dumbbell and sphere-like were synthesized in water/ethanol system under environment-friendly hydrothermal condition. In the synthesis process, the CO 2 in atmosphere was used as the source of carbonate ions and Schiff base was used as shape guiding-agent. The different superstructures of MnCO 3 could be obtained by controlling the hydrothermal temperature, the molar ratio of manganous ions to the Schiff base, or the volume ratio of water to ethanol. A tentative growth mechanism for the generation of MnCO 3 superstructures was proposed based on the rod-dumbbell-sphere model. Furthermore, the MnCO 3 as precursor could be further successfully transferred to Mn 2 O 3 microstructure after heating in the atmosphere at 500 o C, and the morphology of the Mn 2 O 3 was directly determined by that of the MnCO 3 precursor.

  11. Nuclear reactors and fuel storage pools security in France and in Belgium, and the associated reinforcement measures. Press kit

    International Nuclear Information System (INIS)

    Besnard, Manon; Marignac, Yves; Boilley, David; MacKerron, Gordon; Becker, Oda; Lyman, Ed; Zerbib, Jean-Claude; Sotty, Meryl

    2017-10-01

    This report on the security of nuclear reactors and fuel storage pools in France and Belgium draws on the contributions of seven experts from France, Germany, the United Kingdom and the United States - specialists in nuclear safety, security, radioprotection and economics - to bring together the full range of expertise necessary to analyse the problem. Each contributor is only responsible for his/her contribution in the form it was commissioned and separately delivered to Greenpeace France. This report looks at an eminently sensitive subject in an extremely delicate context. Critical analysis and independent expertise must, in line with this, be approached in a way which conciliates two conflicting requirements. The first is that of democratic choice. The security of nuclear facilities in the face of external attacks must be open to public debate. There is no justifiable reason for this issue, a major factor in the assessment of risks related to different industrial and energy options, to be excluded from the democratic process. The public has a fundamental right to information about the risks associated with the operation of nuclear facilities, which includes assessing the risk of external attacks in all its dimensions. It is thus the responsibility of non-institutional experts to contribute to this debate. The second requirement, equally fundamental, is the preservation of public security. Contributing to this debate in a way that enhances the risk of an external attack on nuclear facilities, or, even worse, favours the success of a possible attack by revealing any flaws in the system, is of course out of the question. It is therefore also the responsibility of non-institutional experts to ensure that the protection of such facilities is not impaired by the information they collect or the analyses they produce and make available to the public. It is particularly difficult to strike this balance in the French context, since the authorities responsible for nuclear

  12. Comparative analysis of different joining techniques to improve the passive fit of cobalt-chromium superstructures.

    Science.gov (United States)

    Barbi, Francisco C L; Camarini, Edevaldo T; Silva, Rafael S; Endo, Eliana H; Pereira, Jefferson R

    2012-12-01

    The influence of different joining techniques on passive fit at the interface structure/abutment of cobalt-chromium (Co-Cr) superstructures has not yet been clearly established. The purpose of this study was to compare 3 different techniques of joining Co-Cr superstructures by measuring the resulting marginal misfit in a simulated prosthetic assembly. A specially designed metal model was used for casting, sectioning, joining, and measuring marginal misfit. Forty-five cast bar-type superstructures were fabricated in a Co-Cr alloy and randomly assigned by drawing lots to 3 groups (n=15) according to the joining method used: conventional gas-torch brazing (G-TB), laser welding (LW), and tungsten inert gas welding (TIG). Joined specimens were assembled onto abutment analogs in the metal model with the 1-screw method. The resulting marginal misfit was measured with scanning electron microscopy (SEM) at 3 different points: distal (D), central (C), and mesial (M) along the buccal aspect of both abutments: A (tightened) and B (without screw). The Levene test was used to evaluate variance homogeneity and then the Welsch ANOVA for heteroscedastic data (α=.05). Significant differences were found on abutment A between groups G-TB and LW (P=.013) measured mesially and between groups G-TB and TIG (P=.037) measured centrally. On abutment B, significant differences were found between groups G-TB and LW (Plaser method. Copyright © 2012 The Editorial Council of the Journal of Prosthetic Dentistry. Published by Mosby, Inc. All rights reserved.

  13. System design optimization for stand-alone photovoltaic systems sizing by using superstructure model

    International Nuclear Information System (INIS)

    Azau, M A M; Jaafar, S; Samsudin, K

    2013-01-01

    Although the photovoltaic (PV) systems have been increasingly installed as an alternative and renewable green power generation, the initial set up cost, maintenance cost and equipment mismatch are some of the key issues that slows down the installation in small household. This paper presents the design optimization of stand-alone photovoltaic systems using superstructure model where all possible types of technology of the equipment are captured and life cycle cost analysis is formulated as a mixed integer programming (MIP). A model for investment planning of power generation and long-term decision model are developed in order to help the system engineer to build a cost effective system.

  14. Polymersomes with asymmetric membranes and self-assembled superstructures using pentablock quintopolymers resolved by electron tomography

    KAUST Repository

    Haataja, J. S.

    2018-01-09

    Polystyrene-block-poly(1,4-isoprene)-block-poly(dimethyl siloxane)-block-poly(tert-butyl methacrylate)-block-poly(2-vinyl pyridine), PS-b-PI-b-PDMS-b-PtBMA-b-P2VP, self-assembles in acetone into polymersomes with asymmetric (directional) PI-b-PDMS membranes. The polymersomes, in turn, self-assemble into superstructures. Analogically to supravesicular structures at a smaller length scale, we refer to them as suprapolymersome structures. Electron tomograms are shown to be invaluable in the structural assessment of such complex self-assemblies.

  15. Visualizing the photovoltaic behavior of a type-II p-n heterojunction superstructure

    Energy Technology Data Exchange (ETDEWEB)

    Xing, Juanjuan, E-mail: xingjuanjuan@mail.sic.ac.cn [State Key Laboratory of High Performance Ceramics and Superfine Microstructure, Shanghai Institute of Ceramics, Chinese Academy of Sciences, 1295 Dingxi Road, Shanghai 200050 (China); Electron Microscopy Group, Surface Physics and Structure Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Takeguchi, Masaki [Electron Microscopy Group, Surface Physics and Structure Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Hashimoto, Ayako [Electron Microscopy Group, Surface Physics and Structure Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Global Research Center for Environment and Energy Based on Nanomaterials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Cao, Junyu; Ye, Jinhua [International Center for Materials Nanoarchitectonics (WPI-MANA), 1-1 Namiki, Tsukuba, Ibaraki 305-0044 (Japan)

    2014-04-21

    Photovoltaic behavior of a CaFe{sub 2}O{sub 4}/ZnFe{sub 2}O{sub 4} p-n multi-junction was investigated with electron holography combined with an in situ light irradiation system. Potential profiles of the samples with and without light irradiation were extracted to measure the open circuit photovoltage generated either by the whole heterojunction superstructure or from each p-n junction. Investigation on the variation in the energy band configuration under light irradiation revealed the mechanism involved in the photoelectric effect, with respect to the properties of the heterojunction and its periodic quantum structure.

  16. Time and dose assessment of barge shipment and at-reactor handling of a CASTOR V/21 spent fuel storage cask

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-01

    This report contains the results of a time/motion analysis and a radiation dose assessment made during the receipt from barge transport and the loading of CAst iron cask for Storage and Transport Of Radioactive material (CASTOR) V/21 storage casks with spent nuclear fuel at the Surry Power Station in Virginia during 1987. The study was a cooperative effort between Pacific Northwest Laboratory (PNL) and Virginia Electric and Power Company (Virginia Power), and was funded by the US Department of Energy (DOE) Transportation Program Office. In this study, cask handling activities were tracked at the Surry Power Station, tracing the transfer of the empty spent fuel storage cask from an ocean-going vessel to a barge for river transport through the activities required to place the loaded storage cask at an at-reactor storage location.

  17. Safety aspects of the cleaning and conditioning of radioactive sludge from spent fuel storage pool on 'RA' Research reactor in the Vinca Institute

    International Nuclear Information System (INIS)

    Pavlovic, R; Pavlovic, S.; Plecas, I.

    1999-01-01

    Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some safety aspects and radiation protection measures in the process of the spent fuel storage pool cleaning are presented in this paper

  18. A systems engineering analysis to examine the economic impact for treatment of tritiated water in the Hanford KE-Basin

    International Nuclear Information System (INIS)

    Villegas, A.; Clark, L.; Schmidt, A.

    1995-02-01

    Federal and state agencies have established a Tri-Party Agreement (TPA) to address some key environmental issues faced at the Hanford Site. Under the TPA, the Department of Energy is currently under a consent order to reduce the tritium concentration in the spent fuel storage basin for KE-Reactor from 3.0 μCi/L to 0.3 μCi/L in the KE spent fuel storage basin, starting in 1996. The 100KE and 100KW Area fuel storage basins (K-Basins) at Hanford were built in the early 1950s to receive and provide temporary storage for irradiated fuel from the now shutdown KE and KW production reactors. In 1977, the KE-Basin began to leak at a rate of 13.5 gpm (51 L/min.), but, decreased to 0.03 to 0.05 gpm (0. 13 to 0.19 L/min.) by 1980. In 1993, the leak increased to a rate of 0.42 gpm (1.6 L/min.). This engineering analysis examines the relative costs to reduce the tritium concentration KE-Basin water using a polyphosphazene polymer membrane under development at Pacific Northwest Laboratory. The estimated cost of using the membrane to reduce the tritium concentration is compared to three no-treatment alternatives that include: (1) disposing of the tritium-contaminated water directly to the Columbia River, (2) disposing of the contaminated water to the soil at the on-site Effluent Treatment Facility, and (3) disposing of the contaminated water by evaporation using solar evaporation ponds

  19. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

    2015-07-01

    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 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 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)

  20. Topotactic transformations of superstructures: from thin films to two-dimensional networks to nested two-dimensional networks.

    Science.gov (United States)

    Guo, Chuan Fei; Cao, Sihai; Zhang, Jianming; Tang, Haoying; Guo, Shengming; Tian, Ye; Liu, Qian

    2011-06-01

    Design and synthesis of super-nanostructures is one of the key and prominent topics in nanotechnology. Here we propose a novel methodology for synthesizing complex hierarchical superstructures using sacrificial templates composed of ordered two-dimensional (2D) nanostructures through lattice-directed topotactic transformations. The fabricated superstructures are nested 2D orthogonal Bi(2)S(3) networks composed of nanorods. Further investigation indicates that the lattice matching between the product and sacrificial template is the dominant mechanism for the formation of the superstructures, which agrees well with the simulation results based on an anisotropic nucleation and growth analysis. Our approach may provide a promising way toward a lattice-directed nonlithographic nanofabrication technique for making functional porous nanoarchitectures and electronic devices. © 2011 American Chemical Society

  1. Experimental Investigation of a Self-Sensing Hybrid GFRP-Concrete Bridge Superstructure with Embedded FBG Sensors

    OpenAIRE

    Wang, Yanlei; Li, Yunyu; Ran, Jianghua; Cao, Mingmin

    2012-01-01

    A self-sensing hybrid GFRP-concrete bridge superstructure, which consists of two bridge decks and each bridge deck is comprised of four GFRP box sections combined with a thin layer of concrete in the compression zone, was developed by using eight embedded FBG sensors in the top and bottom flanges of the four GFRP box sections at midspan section of one bridge deck along longitudinal direction, respectively. The proposed self-sensing hybrid bridge superstructure was tested in 4-point loading to...

  2. Comparative analysis of cogeneration power plants optimization based on stochastic method using superstructure and process simulator

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Leonardo Rodrigues de [Instituto Federal do Espirito Santo, Vitoria, ES (Brazil)], E-mail: leoaraujo@ifes.edu.br; Donatelli, Joao Luiz Marcon [Universidade Federal do Espirito Santo (UFES), Vitoria, ES (Brazil)], E-mail: joaoluiz@npd.ufes.br; Silva, Edmar Alino da Cruz [Instituto Tecnologico de Aeronautica (ITA/CTA), Sao Jose dos Campos, SP (Brazil); Azevedo, Joao Luiz F. [Instituto de Aeronautica e Espaco (CTA/IAE/ALA), Sao Jose dos Campos, SP (Brazil)

    2010-07-01

    Thermal systems are essential in facilities such as thermoelectric plants, cogeneration plants, refrigeration systems and air conditioning, among others, in which much of the energy consumed by humanity is processed. In a world with finite natural sources of fuels and growing energy demand, issues related with thermal system design, such as cost estimative, design complexity, environmental protection and optimization are becoming increasingly important. Therefore the need to understand the mechanisms that degrade energy, improve energy sources use, reduce environmental impacts and also reduce project, operation and maintenance costs. In recent years, a consistent development of procedures and techniques for computational design of thermal systems has occurred. In this context, the fundamental objective of this study is a performance comparative analysis of structural and parametric optimization of a cogeneration system using stochastic methods: genetic algorithm and simulated annealing. This research work uses a superstructure, modelled in a process simulator, IPSEpro of SimTech, in which the appropriate design case studied options are included. Accordingly, the cogeneration system optimal configuration is determined as a consequence of the optimization process, restricted within the configuration options included in the superstructure. The optimization routines are written in MsExcel Visual Basic, in order to work perfectly coupled to the simulator process. At the end of the optimization process, the system optimal configuration, given the characteristics of each specific problem, should be defined. (author)

  3. Superstructure formation in PrNi_2Al_3 and ErPd_2Al_3

    International Nuclear Information System (INIS)

    Eustermann, Fabian; Hoffmann, Rolf-Dieter; Janka, Oliver; Oldenburg Univ.

    2017-01-01

    The intermetallic phase ErPd_2Al_3 was obtained by arc-melting of the elements and subsequent annealing for crystal growth. The sample was studied by X-ray diffraction on powders and single crystals. The structure of ErPd_2Al_3 was refined from X-ray diffraction data and revealed a superstructure of PrNi_2Al_3 - a CaCu_5 derivative (P6/m, a=1414.3(1), c=418.87(3) pm wR=0.0820, 1060 F"2 values, 48 variables). The same superstructure was subsequently found for PrNi_2Al_3 (P6/m, a=1407.87(4), c=406.19(2) pm, wR=0.0499, 904 F"2 values, 47 variables). In the crystal structure, the aluminium and transition metal atoms form a polyanionic network according to [T_2Al_3]"δ"-, while rare earth atoms fill cavities within the networks. They are coordinated by six transition metal and twelve aluminum atoms. In contrast to the PrNi_2Al_3 type structure reported so far, two crystallographic independent rare-earth sites are found of which one (1b) is shifted by 1/2 z, causing a distortion in the structure along with a recoloring of the T and Al atoms in the network.

  4. Dynamic cholesteric liquid crystal superstructures photoaligned by one-step polarization holography

    Science.gov (United States)

    Li, Sen-Sen; Shen, Yuan; Chang, Zhen-Ni; Li, Wen-Song; Xu, Yan-Chao; Fan, Xing-Yu; Chen, Lu-Jian

    2017-12-01

    A convenient approach to modulate the fingerprint textures of methyl red (MR) doped cholesteric liquid crystals by asymmetric photoalignment in the green-light waveband is presented, resulting in the generation of voltage-controllable helical superstructures. The interaction between the MR molecules and the incident light polarization determines the initial twisted planar geometry, providing a multivariant control over the stripe directions of fingerprint textures by applying a proper electric field. The key factors for precise manipulation of fingerprint stripes in a predictable and rewritable manner are analyzed theoretically and investigated experimentally, which involves the alignment asymmetry, the ratio of cell gap to natural pitch length, and the chirality of chiral dopant. Dynamic periodic fingerprint textures in shapes of dashed curve and dashed line are further demonstrated by utilizing a facile one-step polarization holography process using two beams with orthogonal circular and orthogonal linear polarizations, respectively. It is believed that the practical approach described in this study would enrich the research contents of self-assembled hierarchical superstructures using soft liquid crystal building blocks.

  5. Preliminary Hazards Analysis of K-Basin Fuel Encapsulation and Storage

    International Nuclear Information System (INIS)

    Strickland, G.C.

    1994-01-01

    This Preliminary Hazards Analysis (PHA) systematically examines the K-Basin facilities and their supporting systems for hazards created by abnormal operating conditions and external events (e.g., earthquakes) which have the potential for causing undesirable consequences to the facility worker, the onsite individual, or the public. The operational activities examined are fuel encapsulation, fuel storage and cooling. Encapsulation of sludges in the basins is not examined. A team of individuals from Westinghouse produced a set of Hazards and Operability (HAZOP) tables documenting their examination of abnormal process conditions in the systems and activities examined in K-Basins. The purpose of this report is to reevaluate and update the HAZOP in the original Preliminary Hazard Analysis of K-Basin Fuel Encapsulation and Storage originally developed in 1991

  6. Test reports for K Basins vertical fuel handling tools

    Energy Technology Data Exchange (ETDEWEB)

    Meling, T.A.

    1995-02-01

    The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

  7. Novel superstructure of the rocksalt type and element distribution in germanium tin antimony tellurides

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Tobias [Department of Chemistry, Ludwig Maximilian University, Butenandtstraße 5-13, 81377 Munich (Germany); Welzmiller, Simon [Institute for Mineralogy, Crystallography and Materials Science, Leipzig University, Scharnhorststraße 20, 04275 Leipzig (Germany); Neudert, Lukas [Department of Chemistry, Ludwig Maximilian University, Butenandtstraße 5-13, 81377 Munich (Germany); Urban, Philipp [Institute for Mineralogy, Crystallography and Materials Science, Leipzig University, Scharnhorststraße 20, 04275 Leipzig (Germany); Fitch, Andy [European Synchrotron Radiation Facility, CS40220, 38043 Grenoble Cedex 9 (France); Oeckler, Oliver, E-mail: oliver.oeckler@gmx.de [Institute for Mineralogy, Crystallography and Materials Science, Leipzig University, Scharnhorststraße 20, 04275 Leipzig (Germany)

    2014-11-15

    A superstructure of the rocksalt-type observed in quenched CVT-grown single crystals of Ge{sub 3.25(7)}Sn{sub 1.10(3)}Sb{sub 1.10(3)}Te{sub 6} was elucidated by X-ray diffraction using fourfold twinned crystals (space group P3{sup ¯}m1, a=4.280(1) Å, c=20.966(3) Å). The structure is built up of distorted rocksalt-type building blocks typical for long-range ordered GST materials and substitution variants thereof. In contrast to those phases, an exclusive ABC-type cubic stacking sequence of the Te-atom layers is present. High-resolution electron microscopy reveals spheroidal domains with this structure (average diameter 25 nm) whose stacking direction is perpendicular to the 〈1 1 1〉 directions of the basic rocksalt-type structure. Additional slab-like domains with a lateral extension up to 1 µm occasionally result in a hierarchical structure motif. Due to the similar electron counts of the elements involved, resonant diffraction was used in order to elucidate the element distribution in rocksalt-type building blocks of the stable layered compound 39R-Ge{sub 3}SnSb{sub 2}Te{sub 7} (R3{sup ¯}m, a=4.24990(4) Å, c=73.4677(9) Å). Sb tends to occupy the atom site close to the van der Waals gaps while Ge concentrates in the center of the building blocks. - Graphical abstract: High-resolution transmission electron micrograph, SAED pattern and reciprocal lattice section of X-ray single crystal data of Ge{sub 3.25}Sn{sub 1.1}Sb{sub 1.1}Te{sub 6} with an 11P-type superstructure of the rocksalt type. - Highlights: • A novel superstructure of the rocksalt-type in the system Ge–Sn–Sb–Te is elucidated. • It combines the cubic stacking of the HT phase with building blocks of the RT phase. • It indicates the ordering mechanism during the phase transition of GST materials. • A hierarchical structure motif is promising with respect to the reduction of κ{sub L}. • Resonant diffraction reveals the element distribution in 39R-Ge{sub 3}SnSb{sub 2}Te{sub 7}.

  8. Study on superstructure in ion co-doped BiFeO3 by using transmission electron microscopy

    Science.gov (United States)

    Pu, Shi-Zhou; Guo, Chao; Li, Mei-Ya; Chen, Zhen-Lian; Zou, Hua-Min

    2015-04-01

    La3+ and V5+ co-doped BiFeO3 ceramics are synthesized by rapid liquid sintering technique. The modulated structure in Bi0.85La0.15Fe0.97V0.03O3 is investigated by using transmission electron microscopy (TEM). Two kinds of superstructures are observed in the samples. One is the component modulated superstructure and twin-domain, which is generated by La3+ ordered substitution for Bi3+ and frequently appears. The chemical composition of the superstructure is explored by x-ray energy dispersive spectroscopy (EDS). The model of the ordered structure is proposed. Simulation based on the model is conducted. The second is the fluorite-type δ-Bi2O3 related superstructure. The relation between the ferroelectric property and the microstructure of the sample is also discussed. Project supported by the National Natural Science Foundation of China (Grant Nos. 51372174, 11074193, and 51132001) and the Fundamental Research Funds for the Central Universities.

  9. A Critical Reassessment of Marxian Base-Superstructure Explanations of the Role of Education in Social Change.

    Science.gov (United States)

    Chun, Kyung-Kap

    1986-01-01

    Examines how four major versions of neo-Marxism (i.e., Hegelian, Phenomenological, Structuralist Marxism, and the Frankfurt School) attempt to overcome the base-superstructure thesis of the political economists of education. Considers the implicit social ontological and epistemological assumptions and the related theory of education of each.…

  10. Extended and quasi-continuous tuning of quantum cascade lasers using superstructure gratings and integrated heaters

    Energy Technology Data Exchange (ETDEWEB)

    Bidaux, Yves, E-mail: yves.bidaux@alpeslasers.ch [Alpes Lasers SA, 1-3 Passsage Max Meuron, CH-2001 Neuchâtel (Switzerland); Institute for Quantum Electronics, ETH-Zurich, CH-8093 Zurich (Switzerland); Bismuto, Alfredo, E-mail: alfredo.bismuto@alpeslasers.ch; Tardy, Camille; Terazzi, Romain; Gresch, Tobias; Blaser, Stéphane; Muller, Antoine [Alpes Lasers SA, 1-3 Passsage Max Meuron, CH-2001 Neuchâtel (Switzerland); Faist, Jerome [Institute for Quantum Electronics, ETH-Zurich, CH-8093 Zurich (Switzerland)

    2015-11-30

    In this work, we demonstrate broad electrical tuning of quantum cascade lasers at 9.25 μm, 8.5 μm, and 4.4 μm in continuous wave operation using Vernier-effect distributed Bragg reflectors based on superstructure gratings. Integrated micro-heaters allow to switch from one Vernier channel to the other, while predictable and mode-hop free tuning can be obtained in each channel modulating the laser current with a side mode suppression ratio as high as 30 dB. The resulting device behaves effectively as a switchable multicolour tunable source. Tuning up to 6.5% of the central wavelength is observed. To prove the importance of the developed devices for high resolution molecular spectroscopy, a N{sub 2}O absorption spectrum has been measured.

  11. Unsupervised Performance Evaluation Strategy for Bridge Superstructure Based on Fuzzy Clustering and Field Data

    Directory of Open Access Journals (Sweden)

    Yubo Jiao

    2013-01-01

    Full Text Available Performance evaluation of a bridge is critical for determining the optimal maintenance strategy. An unsupervised bridge superstructure state assessment method is proposed in this paper based on fuzzy clustering and bridge field measured data. Firstly, the evaluation index system of bridge is constructed. Secondly, a certain number of bridge health monitoring data are selected as clustering samples to obtain the fuzzy similarity matrix and fuzzy equivalent matrix. Finally, different thresholds are selected to form dynamic clustering maps and determine the best classification based on statistic analysis. The clustering result is regarded as a sample base, and the bridge state can be evaluated by calculating the fuzzy nearness between the unknown bridge state data and the sample base. Nanping Bridge in Jilin Province is selected as the engineering project to verify the effectiveness of the proposed method.

  12. Growth and Transfer of Monolithic Horizontal ZnO Nanowire Superstructures onto Flexible Substrates

    KAUST Repository

    Xu, Sheng

    2010-04-28

    A method of fabricating horizontally aligned ZnO nanowire (NW) arrays with full control over the width and length is demonstrated. A cross-sectional view of the NWs by transmission electron microscopy shows a "mushroom-like" structure. Novel monolithic multisegment superstructures are fabricated by making use of the lateral overgrowth. Ultralong horizontal ZnO NWs of an aspect ratio on the order often thousand are also demonstrated. These horizontal NWs are lifted off and transferred onto a flexible polymer substrate, which may have many great applications in horizontal ZnO NW-based nanosensor arrays, light-emitting diodes, optical gratings, integrated circuit interconnects, and high-output-power alternating-current nanogenerators. © 2010 WILEY-VCH Verlag GmbH & Co. KGaA.

  13. Synthesis of noble metal nanoparticles and their superstructures; Darstellung von Edelmetallnanopartikeln und deren Ueberstrukturen

    Energy Technology Data Exchange (ETDEWEB)

    Bigall, Nadja-Carola

    2009-08-18

    A modified synthesis procedure for citrate-stabilized gold nanoparticles in aqueous solution is transferred under application of equal concentrations to the systems silver, platinum, and palladium. The nanoparticles are analyzed by means of absorption spectroscopy and electron microscopy. Ordered superstructures of the noble-metal nanoparticles can be synthesized by infiltration of templates of block-copolymer films with aqueous nanoparticle solution. In dependence on the pre-treatment of the polymer films either two-dimensional periodical arrangements with a periodicity of less than 30 nm or fingerprint-like arrangements with a groove distance in the same order of magnitude. By removal of the polymer one- respectively two-dimensional arrangements of platinum nanowires respectively nanoparticles on a silicon waver arise.

  14. Systematic network synthesis and design: Problem formulation, superstructure generation, data management and solution

    DEFF Research Database (Denmark)

    Quaglia, Alberto; Gargalo, Carina L.; Chairakwongsa, Siwanat

    2015-01-01

    when large problems are considered. In an earlier work, we proposed a computer-aided framework for synthesis and design of process networks. In this contribution, we expand the framework by including methods and tools developed to structure, automate and simplify the mathematical formulation......The developments obtained in recent years in the field of mathematical programming considerably reduced the computational time and resources needed to solve large and complex Mixed Integer Non Linear Programming (MINLP) problems. Nevertheless, the application of these methods in industrial practice...... is still limited by the complexity associated with the mathematical formulation of some problems. In particular, the tasks of design space definition and representation as superstructure, as well as the data collection, validation and handling may become too complex and cumbersome to execute, especially...

  15. Superstructure of self-aligned hexagonal GaN nanorods formed on nitrided Si(111) surface

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Praveen; Tuteja, Mohit; Kesaria, Manoj; Waghmare, U. V.; Shivaprasad, S. M. [Jawaharlal Nehru Centre for Advanced Scientific Research, Jakkur, Bangalore 560 064 (India)

    2012-09-24

    We present here the spontaneous formation of catalyst-free, self-aligned crystalline (wurtzite) nanorods on Si(111) surfaces modified by surface nitridation. Nanorods grown by molecular beam epitaxy on bare Si(111) and non-stoichiometric silicon nitride interface are found to be single crystalline but disoriented. Those grown on single crystalline Si{sub 3}N{sub 4} intermediate layer are highly dense c-oriented hexagonal shaped nanorods. The morphology and the self-assembly of the nanorods shows an ordered epitaxial hexagonal superstructure, suggesting that they are nucleated at screw dislocations at the interface and grow spirally in the c-direction. The aligned nanorod assembly shows high-quality structural and optical emission properties.

  16. Experimental and theoretical study of the operation of longitudinal sidewalls in reinforced concrete superstructures

    Science.gov (United States)

    Bokarev, Sergey; Efimov, Stefan

    2017-11-01

    In the present article, we report results of a laboratory experimental study of reinforced-concrete superstructure fragments with extended longitudinal boxing sidewalls. The study included laboratory tests and numerical experiments performed using a finite element model of fragments generated in specialized simulation software. The distribution of horizontal pressures over the height of longitudinal boxing sidewalls, the dependence of the bending moment in the root section of the sidewalls on the ballast-bed thickness under sleepers, and some other characteristics were examined. The results of the experimental and theoretical study have shown a good convergence of obtained data to each other and to the data obtained by an improved engineering calculation procedure proposed by the present authors.

  17. Prevention of Cutaneous Tissue Contracture During Removal of Craniofacial Implant Superstructures for CT and MRI Studies

    Directory of Open Access Journals (Sweden)

    Maureen Sullivan

    2010-04-01

    Full Text Available Objectives: Head and neck cancer patients who have lost facial parts following surgical intervention frequently require craniofacial implant retained facial prostheses for restoration. Many craniofacial implant patients require computed tomography and magnetic resonance imaging scans as part of their long-term follow-up care. Consequently removal of implant superstructures and peri-abutment tissue management is required for those studies. The purpose of the present paper was to describe a method for eliminating cranial imaging artifacts in patients with craniofacial implants.Material and Methods: Three patients wearing extraoral implant retained facial prostheses needing either computed tomography or magnetic resonance imaging studies were discussed. Peri-implant soft tissues contracture after removal of percutaneous craniofacial implant abutments during computed tomography and magnetic resonance imaging studies was prevented using a method proposed by authors. The procedure involves temporary removal of the supra-implant components prior to imaging and filling of the tissue openings with polyvinyl siloxane dental impression material.Results: Immediately after filling of the tissue openings with polyvinyl siloxane dental impression material patients were sent for the imaging studies, and were asked to return for removal of the silicone plugs and reconnection of all superstructure hardware after imaging procedures were complete. The silicone plugs were easily removed with a dental explorer. The percutaneous abutments were immediately replaced and screwed into the implants which were at the bone level.Conclusions: Presented herein method eliminates the source of artifacts and prevents contracture of percutaneous tissues upon removal of the implant abutments during imaging.

  18. Enhanced fullerene–Au(111 coupling in (2√3 × 2√3R30° superstructures with intermolecular interactions

    Directory of Open Access Journals (Sweden)

    Michael Paßens

    2015-06-01

    Full Text Available Disordered and uniform (2√3 × 2√3R30° superstructures of fullerenes on the Au(111 surface have been studied using scanning tunneling microscopy and spectroscopy. It is shown that the deposition and growth process of a fullerene monolayer on the Au(111 surface determine the resulting superstructure. The supply of thermal energy is of importance for the activation of a Au vacancy forming process and thus, one criterion for the selection of the respective superstructure. However, here it is depicted that a vacancy–adatom pair can be formed even at room temperature. This latter process results in C60 molecules that appear slightly more bright in scanning tunnelling microscopy images and are identified in disordered (2√3 x 2√3R30° superstructures based on a detailed structure analysis. In addition, these slightly more bright C60 molecules form uniform (2√3 x 2√3R30° superstructures, which exhibit intermolecular interactions, likely mediated by Au adatoms. Thus, vacancy–adatom pairs forming at room temperature directly affect the resulting C60 superstructure. Differential conductivity spectra reveal a lifting of the degeneracy of the LUMO and LUMO+1 orbitals in the uniform (2√3 x 2√3R30° superstructure and in addition, hybrid fullerene–Au(111 surface states suggest partly covalent interactions.

  19. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    International Nuclear Information System (INIS)

    1978-06-01

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions

  20. Risk-based determination of design pressure of LNG fuel storage tanks based on dynamic process simulation combined with Monte Carlo method

    International Nuclear Information System (INIS)

    Noh, Yeelyong; Chang, Kwangpil; Seo, Yutaek; Chang, Daejun

    2014-01-01

    This study proposes a new methodology that combines dynamic process simulation (DPS) and Monte Carlo simulation (MCS) to determine the design pressure of fuel storage tanks on LNG-fueled ships. Because the pressure of such tanks varies with time, DPS is employed to predict the pressure profile. Though equipment failure and subsequent repair affect transient pressure development, it is difficult to implement these features directly in the process simulation due to the randomness of the failure. To predict the pressure behavior realistically, MCS is combined with DPS. In MCS, discrete events are generated to create a lifetime scenario for a system. The combination of MCS with long-term DPS reveals the frequency of the exceedance pressure. The exceedance curve of the pressure provides risk-based information for determining the design pressure based on risk acceptance criteria, which may vary with different points of view. - Highlights: • The realistic operation scenario of the LNG FGS system is estimated by MCS. • In repeated MCS trials, the availability of the FGS system is evaluated. • The realistic pressure profile is obtained by the proposed methodology. • The exceedance curve provides risk-based information for determining design pressure

  1. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  2. Structural evaluation of spent nuclear fuel storage facilities under aircraft crash impact (2). Horizontal impact test onto reduced scale metal cask due to aircraft engine missile

    International Nuclear Information System (INIS)

    Namba, Kosuke; Shirai, Koji; Saegusa, Toshiari

    2009-01-01

    In this study, to confirm the sealing performance of a metal cask subjected to impact force due to possible commercial aircraft crash against a spent fuel storage facility, the horizontal impact test was carried out. In the test, an aircraft engine missile with a speed of 57.3 m/s attacked the reduced scale metal cask containing helium gas, which stands vertically. Then the leak rate and sliding displacement of the lid were measured. The leak rate increased rapidly and reached to 4.0 x 10 -6 Pa·m 3 /sec. After that, the leak rate decreased slowly and converged to 1.0x10 -6 Pa·m 3 /sec after 20 hours from the impact test. The leak rate of a full scale cask was evaluated using that of reduced scale cask obtained by the test. Then the leak rate of the full scale cask was 3.5x10 -5 Pa·m 3 /sec. This result showed that the sealing performance of the full scale metal cask would not be affected immediately by the horizontal impact of the aircraft engine with a speed of 57.3 m/s. (author)

  3. A model of the dose rate calculation for a spent fuel storage structure by Monte Carlo method using the modulated code system SCALE 4.4a

    International Nuclear Information System (INIS)

    Pantazi, D.; Mateescu, S.; Stanciu, M.; Mete, M.

    2001-01-01

    The modulated code system SCALE is used to perform a standardized shielding analysis for any facility containing spent fuel: handling devices, transport cask, intermediate and final storage facility. The neutron and gamma sources as well as the dose rates can be obtained using either discrete-ordinates or Monte Carlo methods. The shielding analysis control modules (SAS1, SAS2H and SAS4) provide a general procedure for cross-section preparation, fuel depletion/decay calculation and general onedimensional or multi-dimensional shielding analysis. The module SAS4 used in the analysis presented in this paper, is a three-dimensional Monte Carlo shielding analysis module, which uses an automated biasing procedure specialized for a nuclear fuel transport or storage container. The Spent Fuel Interim Storage Facility in our country is projected to be a parallelepiped concrete monolithic module, consisting of an external reinforced concrete structure with vertical storage cylinders (pits) arranged in a rectangular array. A pit is filled with sealed cylindrical baskets of stainless steel arranged in a stack, and with each basket containing spent fuel bundles in vertical position. The pit is closed with a concrete plug. The cylindrical geometry model is used in the shielding evaluation for a spent fuel storage structure (pit), and only the active parts of the superposed bundles is considered. The dose rates have been calculated in both the axial and radial directions using SAS4.(author)

  4. Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1987-01-01

    The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

  5. Simultaneous measurement of temperature and tensile loading using superstructure FBGs developed by laser direct writing of periodic on-fiber metallic films

    International Nuclear Information System (INIS)

    Alemohammad, Hamidreza; Toyserkani, Ehsan

    2009-01-01

    This paper addresses the development of superstructure fiber Bragg gratings (FBGs) by laser-assisted direct writing of on-fiber metallic films. A novel laser direct write method is characterized to fabricate periodic films of silver nanoparticles on the non-planar surface of as-fabricated FBGs. Silver films with a thickness of 9 µm are fabricated around a Bragg grating optical fiber. The performance of the superstructure FBG is studied by applying temperature and tensile stress on the fiber. An opto-mechanical model is also developed to predict the optical response of the synthesized superstructure FBG under thermal and structural loadings. The results show that the reflectivity of sidebands in the reflection spectrum can be tuned up to 20% and 37% under thermal and structural loadings, respectively. In addition, the developed superstructure FBG is used for simultaneous measurement of force and temperature to eliminate the inherent limitation of regular FBGs in multi-parameter sensing

  6. Presenting a Multi-level Superstructure Optimization Approach for Mechatronic System Design

    DEFF Research Database (Denmark)

    Pedersen, Henrik C.; Andersen, Torben Ole; Bech, Michael Møller

    2010-01-01

    Synergism and integration in the design process is what sets apart a Mechatronic System from a traditional, multidisciplinary system. However the typical design approach has been to divide the design problem into sub problems for each technology area (mechanics, electronics and control) and descr......Synergism and integration in the design process is what sets apart a Mechatronic System from a traditional, multidisciplinary system. However the typical design approach has been to divide the design problem into sub problems for each technology area (mechanics, electronics and control......) and describe the interface between the technologies, whereas the lack of well-established, systematic engineering methods to form the basic set-off in analysis and design of complete mechatronic systems has been obvious. The focus of the current paper is therefore to present an integrated design approach...... for mechatronic system design, utilizing a multi-level superstructure optimization based approach. Finally two design examples are presented and the possibilities and limitations of the approach are outlined....

  7. Solution Construction of Multigeometry Nanoparticles and Multicompartment Superstructures from Block Copolymer Mixtures

    Science.gov (United States)

    Zhu, Jiahua; Zhang, Shiyi; Wooley, Karen; Pochan, Darrin

    2013-03-01

    Novel soft objects with both compositional and geometric complexity at nanoscale have been constructed through solution supramolecular assembly from block copolymer mixtures due to their non-ergodic character. The mixture is composed of two block copolymers with distinctive hydrophobic blocks but the same poly(acrylic acid) hydrophilic block. First, multigeometry nanoparticles, due to segregation of unlike block copolymer molecules into multiple subdomains trapped within the same micelle-like structures, have been assembled in tetrahydrofuran/water solution. Through carefully designed molecular architecture, mixing ratio and pathway kinetics, both size and shape of subdomains can be controlled to produce a novel class of multigeometry nanoparticles, including sphere-sphere, sphere-cylinder, cylinder-cylinder, cylinder-disk, and sphere-disk hybrid nanoparticles. Second, hierarchical multicompartment superstructures including particle chains, rings and other nano to micro cluster formations, have been built up from pre-formed multigeometry nanoparticles by taking advantage of their surface anisotropy and the controlled particle-particle association. The interparticle association can be achieved via either covalent or non-covalent bindings due to different post-polymerization chemical modifications with hydroxyethyl acrylate or crown ether functionalities, respectively.

  8. High density fuel storage racks

    International Nuclear Information System (INIS)

    Groves, M.D.

    1978-01-01

    An apparatus is described for the safe and compact storage of nuclear fuel assemblies in an array of discrete open-ended neutron absorbing shields for which the theoretical minimum safe separation distance and cell pitch are known. Open-ended stainless steel end fittings are welded to each end of each shield and the end fittings are welded to each other in side-by-side relation, thereby reducing the cell pitch tolerance due to fabrication uncertainties. In addition, a multiplicity of ridges on the sides of each shield having a height equal to one half the theoretical minimum safe separation distance further reduce shield bowing tolerances. The net tolerance reduction permits a significant increase in the number of fuel assemblies that can be safely contained in a storage area of fixed size

  9. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  10. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

    Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  11. Spent nuclear fuel storage vessel

    International Nuclear Information System (INIS)

    Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

    1997-01-01

    Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)

  12. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

    1987-09-01

    Historical inventories of spent fuel and utility estimates of future discharges from US commercial nuclear reactors are presented through the year 2005. The ultimate needs for additional storage capacity are estimated. These estimtes are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December, 1986, and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to the DOE Energy Information Administration (EIA) through the 1987 RW-859 data survey. 14 refs., 4 figs., 9 tabs

  13. Bone-repair properties of biodegradable hydroxyapatite nano-rod superstructures

    Science.gov (United States)

    D'Elía, Noelia L.; Mathieu, Colleen; Hoemann, Caroline D.; Laiuppa, Juan A.; Santillán, Graciela E.; Messina, Paula V.

    2015-11-01

    Nano-hydroxyapatite (nano-HAp) materials show an analogous chemical composition to the biogenic mineral components of calcified tissues and depending on their topography they may mimic the specific arrangement of the crystals in bone. In this work, we have evaluated the potential of four synthesized nano-HAp superstructures for the in vitro conditions of bone-repair. Experiments are underway to investigate the effects of the material microstructure, surface roughness and hydrophilicity on their osseo-integration, osteo-conduction and osteo-induction abilities. Materials were tested in the presence of both, rat primary osteoblasts and rabbit mesenchymal stem cells. The following aspects are discussed: (i) cytotoxicity and material degradation; (ii) rat osteoblast spreading, proliferation and differentiation; and (iii) rabbit mesenchymal stem cell adhesion on nano-HAp and nano-HAp/collagen type I coatings. We effectively prepared a material based on biomimetic HAp nano-rods displaying the appropriate surface topography, hydrophilicity and degradation properties to induce the in vitro desired cellular responses for bone bonding and healing. Cells seeded on the selected material readily attached, proliferated and differentiated, as confirmed by cell viability, mitochondrial metabolic activity, alkaline phosphatase (ALP) activity and cytoskeletal integrity analysis by immunofluorescence localization of alpha-smooth muscle actin (α-SMA) protein. These results highlight the influence of material's surface characteristics to determine their tissue regeneration potential and their future use in engineering osteogenic scaffolds for orthopedic implants.Nano-hydroxyapatite (nano-HAp) materials show an analogous chemical composition to the biogenic mineral components of calcified tissues and depending on their topography they may mimic the specific arrangement of the crystals in bone. In this work, we have evaluated the potential of four synthesized nano-HAp superstructures

  14. Base, Superstructure and the Irish Property Crash—Towards a Crisis Theory of Communications

    Directory of Open Access Journals (Sweden)

    Henry Silke

    2015-09-01

    Full Text Available Since the onset of the “great recession” there have been key debates around various aspects of crisis theory, most notably around the areas of the rate of profit (Brenner 2009; Kliman 2012, under-consumption/overproduction (Clarke 1990a, 442–467 and fiancialisation (Duménil and Lévy 2004. This paper maintains that communications and the media are key though non-deterministic elements of the contemporary market system, and proposes a move towards a crisis theory of communications. This research reflects the Marxist concept of base and superstructure, beyond a perceived notion of economic determinism, but rather as a dialectical relationship between various superstructures, in this case the state and the media, and the economic base including the various aspects of class power inherent within. The mass media, advertising, and ICT play an increasingly important role in both market systems and capitalist crises. This role directly impinges on the dissemination of information to market actors as well as the reflexive and dialectical nature of the processes by which actors respond to market information. Further, the media serve as an ideological apparatus, resource or arena, which acts to naturalise the market through what this research describes as a market orientated framing mechanism (Preston and Silke 2011b. Peter Thompson (2003; 2013, 208–227 contends that communication is an integral and reflexive part of the contemporary market system. As he puts it, there is a complex relationship between the producers and distributors of economic information, and those who use that information to make decisions about investment and trade. Many studies point to the convergence of flows of information such as those on 24-hour news channels, business channels and Internet blogs and sites with market activity itself. For Wayne Hope, (2010, 649–669 information broadcast on such media by bankers, stockbrokers and traders themselves tends to be self

  15. The importance of DNA superstructure units for the understanding of the radiation action mechanism

    International Nuclear Information System (INIS)

    Regel, K.

    1985-04-01

    A molecular radiation action model is presented. It relates the physical parameters of the radiation interaction in tissue and of the DNA structure in mammalian cells to their dose survival curves. Using this model it is possible to explain many of the radiation effects in cells, including such ones which were not clearly understood as yet. Both the kind of the basic parameters and the 'efficiency' of the model suggest that it describes real properties of mammalian cells. However, in finding out the radiation action mechanism we had to fill up two gaps in our knowledge concerning the radiation action in organisms. The first gap is characterized by the question: Are there any DNA structures (sites) in mammalian cells on the basis of which a radiation action model can be established which is valid in all the cell cycle stages. This question is answered by comparisons of the magnitude of DNA parameters measured in suitable experiments with those calculated from a hypothetical model of DNA organization in mammalian cells. The second gap in knowledge is filled up by testing the hypothesis that certain patterns of double-strand breaks (DSBs) in the membrane attached superstructure units (MASSUs) of a cell cause its inactivation. The dependence of the dose survival curves on the cell cycle can be explained in the following way: Dose survival curves of G1, G2 and mitotic cells are changed because of the cyclically altering volume of the MASSU compartments. Its change during the S stage is mainly determined by the growing fraction of replicated MASSUs. The high radiation resistance of late S cells probably results from the ability of mammalian cells to establish one intact sister genome from both sister genomes containing heavily damaged MASSUs joint in the attachment points. This ability is explained by the interference of DSB repair, sister chromatid exchange and DNA degradation. (author)

  16. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Raymond, R. E.; Evans, K. M.

    2012-01-01

    CH2M Hill Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material)

  17. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    International Nuclear Information System (INIS)

    Mays, Gary T.; Belles, Randy; Cetiner, Mustafa Sacit; Howard, Rob L.; Liu, Cheng; Mueller, Don; Omitaomu, Olufemi A.; Peterson, Steven K.; Scaglione, John M.

    2012-01-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  18. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  19. FY 1998 annual report on the fuel/storage working group. 19th R and D activity report; 1998 nendo nenryo chozo gijutsu bunkakai. Dai 19 kai jigyo hokokukai

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-09-01

    Summarized herein are the FY 1998 R and D activities by the fuel/storage working group, extracted from the 19th R and D activity report by NEDO. Mr. Kadoi, a NEDO's director, outlined the R and D activities for development of techniques for fuel cell power generation, development of decentralized cell power storage systems, demonstration of new load-leveling techniques, development of techniques for superlow-loss type power system components, development of techniques for ceramic gas turbines, pioneer R and D for MGC superhigh-efficiency turbine system techniques, demonstration tests for establishing centralized load controlling systems, development of techniques for network systems utilizing wide-area energy, development of techniques for application of superconductor-aided power systems, R and D for power storage aided by a high-temperature, superconducting flywheel, R and D for basic techniques for applying superconductors, and pioneer R and D for basic techniques for AC superconducting power equipment. The individual NEDO's researchers reported their results, which include outline of the fuel/storage technique development group, and current statuses of fuel cell technique development, new cell-aided power storage technique development, and superconductor technique development. (NEDO)

  20. Failure behavior of concrete pile and super-structure dynamic response as a result of soil liquefaction during earthquake

    Science.gov (United States)

    Kaneda, Shogo; Hayashi, Kazuhiro; Hachimori, Wataru; Tamura, Shuji; Saito, Taiki

    2017-10-01

    In past earthquake disasters, numerous building structure piles were damaged by soil liquefaction occurring during the earthquake. Damage to these piles, because they are underground, is difficult to find. The authors aim to develop a monitoring method of pile damage based on superstructure dynamic response. This paper investigated the relationship between the damage of large cross section cementitious piles and the dynamic response of the super structure using a centrifuge test apparatus. A dynamic specimen used simple cross section pile models consisting of aluminum rod and mortar, a saturated soil (Toyoura sand) of a relative density of 40% and a super structure model of a natural period of 0.63sec. In the shaking table test under a 50G field (length scale of 1/50), excitation was a total of 3 motions scaled from the Rinkai wave at different amplitudes. The maximum acceleration of each of the excitations was 602gal, 336gal and 299gal. The centrifuge test demonstrated the liquefaction of saturated soil and the failure behavior of piles. In the test result, the damage of piles affected the predominant period of acceleration response spectrum on the footing of the superstructure.

  1. The monoclinic superstructure of the M2Pt6Al15 series (M=Ca, Sc, Y, La, Lu)

    International Nuclear Information System (INIS)

    Radzieowski, Mathis; Stegemann, Frank; Hoffmann, Rolf-Dieter; Janka, Oliver; Oldenburg Univ.

    2017-01-01

    The five ternary intermetallic compounds M 2 Pt 6 Al 15 (M=Ca, Sc, Y, La, Lu) were prepared from the elements by arc-melting. The crystal structure was determined via single crystal X-ray diffraction. The title compounds crystallize in a superstructure of the RE 0.67 Pt 2 Al 5 type structure (P6 3 /mmc) in the monoclinic crystal system with space group P12 1 /m1 (Sc 2 Pt 6 Al 15 : a=734.19(2), b=1628.96(10), c=734.19(2) pm, β=119.999(3) ; wR=0.0356, 3034 F 2 values, 68 variables). The superstructure can be derived by the superspace formalism using (3+2)D or (3+1)D interpretations of the diffraction data. The structural relation to the subcell structure is discussed on the basis of a group-subgroup scheme. In the crystal structure strongly bonded [Pt 2 Al 4 ] δ- slabs are alternatingly stacked with ordered layers containing M atoms and Al 3 triangles.

  2. Self-organization of nickel nanoparticles dispersed in acetone: From separate nanoparticles to three-dimensional superstructures

    Directory of Open Access Journals (Sweden)

    I. Hernández-Pérez

    2017-02-01

    Full Text Available Sonochemical synthesis of monodisperse nickel nanoparticles (Ni-NPs by reduction of Ni acetylacetonate in the presence of polyvinylpyrrolidone stabilizer is reported. The Ni-NPs size is readily controlled to 5 nanometer diameter with a standard deviation of less than 5%. The as-prepared Ni-NPs sample was dispersed in acetone, for 4 weeks. For structural analysis was not applied to a magnetic field or heat treatment as key methods to direct the assembly. The transition from separate Ni-NPs into self-organization of three dimensions (3D superstructures was studied by electron microscopy. Experimental analysis suggests that the translation and rotation movement of the Ni-NPs are governed by magnetic frustration which promotes the formation of different geometric arrangements in two dimensions (2D. The formation of 3D superstructures is confirmed from scanning electron microscopy revealing a layered domain that consists of staking of several monolayers having multiple well-defined supercrystalline domains, enabling their use for optical, electronic and sensor applications.

  3. Analysis of connection element classes and locations and of some structural requirements for the mounting of different superstructure types on transport vehicles

    Directory of Open Access Journals (Sweden)

    Zoran Đ. Majkić

    2011-04-01

    Full Text Available The paper presents the basic requirements for transport vehicles. A special request regarding the adaptation of transport vehicles for the transport of various types of cargo was taken into consideration. Superstructures and the situation arising after mounting superstructures on wheeled transport vehicles were analyzed and the following was described: console coupling, stirrups, simplex elastic coupling, two-way elastic and rigid connection elements. Vehicle torsional elasticity is provided by a proper choice of the type of connection between the superstructure and the vehicle chassis. Applying the instructions of vehicle manufacturers for using appropriate connections between the truck superstructure and the vehicle chassis provides positive torsional elasticity of the vehicle. The paper gives the general recommendations of the Volvo, Mercedes and Renault transport vehicle producers for the use of particular connection types of locations as well as structural requirements for the mounting of concrete mixers, tippers and truck tanks on their vehicles. Introduction Achieving a high level of transport effectiveness depends on a number of factors. One of the most important ones is the possibility to increase the payload share in the gross vehicle weight. This share depends on the net vehicle weight, a method of coupling the truck superstructure with the chassis frame as well as on the truck superstructure construction. Realization of this requirement is of significant importance, particularly for large business systems since it results in the reduction of number of necessary vehicles, more economic fleet maintenance and the fleet capacity increase. It is also relatively easy to adapt the vehicle for the transportation of other loads, depending on user's current needs. The adaptation is correctly performed if manufacturer's recommendations are followed during the mounting of the superstructure on the chassis. This paper gives the analysis of the

  4. Ag on Ge(111): 2D x-ray structure analysis of the #sq root#3 x #sq root#3 superstructure

    DEFF Research Database (Denmark)

    Dornisch, D.; Moritz, W.; Schulz, H.

    1992-01-01

    We have studied the Ag/Ge(111) square-root 3 x square-root 3 superstructure by grazing-incidence X-ray diffraction. In our structural analysis we find striking similarities to the geometry of Au on Si(111). The Ag atoms form trimer clusters with an Ag-Ag distance of 2.94 +/- 0.04 angstrom...

  5. A novel high-temperature commensurate superstructure in a natural bariopyrochlore: A structural study by means of a multiphase crystal structure refinement

    Czech Academy of Sciences Publication Activity Database

    Bindi, L.; Petříček, Václav; Withers, R. L.; Zoppi, M.; Bonazzi, P.

    2006-01-01

    Roč. 179, - (2006), s. 716-725 ISSN 0022-4596 R&D Projects: GA ČR(CZ) GA202/03/0430 Institutional research plan: CEZ:AV0Z10100521 Keywords : pyrochlore * superstructure * X-ray data * multiphase structure refinement * TEM study Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.107, year: 2006

  6. Optomagnetic Detection of MicroRNA Based on Duplex-Specific Nuclease-Assisted Target Recycling and Multilayer Core-Satellite Magnetic Superstructures

    DEFF Research Database (Denmark)

    Tian, Bo; Ma, Jing; Qiu, Zhen

    2017-01-01

    -efficiency, and potential for bioresponsive multiplexing. Herein, we demonstrate a sensitive and rapid miRNA detection method based on optomagnetic read-out, duplex-specific nuclease (DSN)-assisted target recycling, and the use of multilayer core-satellite magnetic superstructures. Triggered by the presence of target mi...

  7. A novel high-temperature commensurate superstructure in a natural bariopyrochlore: A structural study by means of a multiphase crystal structure refinement

    International Nuclear Information System (INIS)

    Bindi, L.; Petricek, V.; Withers, R.L.; Zoppi, M.; Bonazzi, P.

    2006-01-01

    Additional X-ray diffraction effects yielding an eightfold commensurate superstructure [a=20.974(5)A] of the ideal pyrochlore structure were observed after annealing at 873K of a thallium-doped bariopyrochlore single crystal. Electron diffraction indicated the coexistence of two cubic phases, the pyrochlore structure and a new F-centred, cubic phase. The superstructure was solved and refined in the space group F4-bar 3m. The two phases were combined together and refined as independently diffracting to R=0.0628. The resulting unit-cell content is (A,-bar ) 20 Nb 16 Ti 2 O 53 (Z=8), with A=Ba, Tl, Ce, Th. For some atomic positions of the superstructure, third- and fourth-order anharmonic ADP's were used to account for the specific density shape having a continuous character as typical for ionic conductors. There are three distinct clusters in the superstructure, leading to a new structure type no longer strictly of pyrochlore-structure type

  8. N-Basin water removal plan

    International Nuclear Information System (INIS)

    Nellesen, A.L.

    1997-07-01

    This ALARA review provides a description of the engineering and administrative controls used to manage personnel exposure, control contamination levels, and airborne radioactivity concentrations, while removing water and stabilizing surfaces in the 105-N Fuel Storage Building

  9. Topotactic epitaxy of SrTiO3 mesocrystal superstructures with anisotropic construction for efficient overall water splitting

    International Nuclear Information System (INIS)

    Zhang, Peng; Fujitsuka, Mamoru; Majima, Tetsuro; Ochi, Tomoya; Kobori, Yasuhiro; Tachikawa, Takashi

    2017-01-01

    The higher-order structures of semiconductor-based photocatalysts play crucial roles in their physicochemical properties for efficient light-to-energy conversion. A novel perovskite SrTiO 3 mesocrystal superstructure with well-defined orientation of assembled cubic nanocrystals was synthesized by topotactic epitaxy from TiO 2 mesocrystals through a facile hydrothermal treatment. The SrTiO 3 mesocrystal exhibits three times the efficiency for the hydrogen evolution of conventional disordered systems in alkaline aqueous solution. It also exhibits a high quantum yield of 6.7 % at 360 nm in overall water splitting and even good durability up to 1 day. Temporal and spatial spectroscopic observations revealed that the synergy of the efficient electron flow along the internal nanocube network and efficient collection at the larger external cubes produces remarkably long-lived charges for enhanced photocatalysis. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  10. Topotactic epitaxy of SrTiO{sub 3} mesocrystal superstructures with anisotropic construction for efficient overall water splitting

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Peng; Fujitsuka, Mamoru; Majima, Tetsuro [The Institute of Scientific and Industrial Research (SANKEN), Osaka University (Japan); Ochi, Tomoya [Department of Chemistry, Graduate School of Science, Kobe University (Japan); Kobori, Yasuhiro [Department of Chemistry, Graduate School of Science, Kobe University (Japan); Molecular Photoscience Research Center, Kobe University (Japan); Tachikawa, Takashi [Department of Chemistry, Graduate School of Science, Kobe University (Japan); Molecular Photoscience Research Center, Kobe University (Japan); PRESTO, Science and Technology Agency (JST), Saitama (Japan)

    2017-05-02

    The higher-order structures of semiconductor-based photocatalysts play crucial roles in their physicochemical properties for efficient light-to-energy conversion. A novel perovskite SrTiO{sub 3} mesocrystal superstructure with well-defined orientation of assembled cubic nanocrystals was synthesized by topotactic epitaxy from TiO{sub 2} mesocrystals through a facile hydrothermal treatment. The SrTiO{sub 3} mesocrystal exhibits three times the efficiency for the hydrogen evolution of conventional disordered systems in alkaline aqueous solution. It also exhibits a high quantum yield of 6.7 % at 360 nm in overall water splitting and even good durability up to 1 day. Temporal and spatial spectroscopic observations revealed that the synergy of the efficient electron flow along the internal nanocube network and efficient collection at the larger external cubes produces remarkably long-lived charges for enhanced photocatalysis. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. The effect of heat treatments applied to superstructure porcelain on the mechanical properties and microstructure of lithium disilicate glass ceramics.

    Science.gov (United States)

    Özdemir, Hatice; Özdoğan, Alper

    2018-01-30

    The aim of this study was to investigate that heat treatments with different numbers applied to superstructure porcelain whether effects microstructure and mechanical properties of lithium disilicate ceramic (LDC). Eighty disc-shaped specimens were fabricated from IPS e.max Press. Specimens were fired at heating values of porcelain in different numbers and divided four groups (n=5). Initial Vickers hardness were measured and X-ray diffraction (XRD) analysis was performed. Different surface treatment were applied and then Vickers hardness, surface roughness and environmental scanning electron microscopy (ESEM) analysis were performed. Data were analyzed with Varyans analysis and Tukey HSD test (α=0.05). Initial hardness among groups was no significant different (p>0.05), but hardness and surface roughness after surface treatments were significant different (pmicrostructure of LDC. Increasing firing numbers and surface treatments effect the microstructure of LDC.

  12. Hierarchical assembly of inorganic nanostructure building blocks to octahedral superstructures-a true template-free self-assembly

    International Nuclear Information System (INIS)

    Kuchibhatla, Satyanarayana V N T; Karakoti, Ajay S; Seal, Sudipta

    2007-01-01

    A room temperature, template-free, wet chemical synthesis of ceria nanoparticles and their long term ageing characteristics are reported. High resolution transmission electron microscopy and UV-visible spectroscopy techniques are used to observe the variation in size, structure and oxidation state, respectively as a function of time. The morphology variation and the hierarchical assembly (octahedral superstructure) of nanostructures are imputed to the inherent structural aspects of cerium oxide. It is hypothesized that the 3-5 nm individual building blocks will undergo an intra-agglomerate re-orientation to attain the low energy configuration. This communication also emphasizes the need for long term ageing studies of nanomaterials in various solvents for multiple functionalities

  13. A volatile-solvent gas fiber sensor based on polyaniline film coated on superstructure fiber Bragg gratings

    International Nuclear Information System (INIS)

    Ai, L; Chen, T C; Su, W K; Mau, J C; Liu, W F

    2008-01-01

    A fiber sensor based on a polyaniline (PANI) film that is coated on the surface of an etched superstructure fiber grating to detect volatile solvent vapors is experimentally demonstrated. This sensing mechanism is based on the interaction of the testing gas with the polyaniline coating film, which changes the film index, resulting in a shift in the Bragg wavelength. The sensitivity of this sensor to ammonia (NH 3 ) gas is about 0.073 pm ppm −1 , which depends on the optical characteristics of the fiber grating, the diameter of the fiber cladding and the constituents of the sensing film. Methanol concentrations can also be measured using this sensing scheme. The sensitivity of this sensor must be improved to provide a simple, reliable, repeatable and non-destructive method for sensing various chemical gases. (technical design note)

  14. Superstructure of the superconductor BI2Sr2CaCu2O8 by high resolution electron microscopy

    International Nuclear Information System (INIS)

    Hewat, E.A.

    1988-01-01

    The superstructure of the high Tc superconducting oxide BI 2S r 2C aCu 2 0 8 has been studied by high resolution electron microscopy. Waves of distortion along the b-axis give an incommensurate superlattice slightly larger or smaller than 5√2a p (a p = perovskite unit cell). The building blocks along the b-axis are 4, 5 and 6 times √2a p /2. The incommensurate superlattice is composed of approximately periodic combinations of these building blocks. The symmetry of three major projections are P gm (or possibly P gg ), C mm and C mm for the ideal superlattice with b=5√ 2a p . These projections correspond to the space groups Pcnn and Pmnn respectively

  15. A bio-inspired N-doped porous carbon electrocatalyst with hierarchical superstructure for efficient oxygen reduction reaction

    Science.gov (United States)

    Miao, Yue-E.; Yan, Jiajie; Ouyang, Yue; Lu, Hengyi; Lai, Feili; Wu, Yue; Liu, Tianxi

    2018-06-01

    The bio-inspired hierarchical "grape cluster" superstructure provides an effective integration of one-dimensional carbon nanofibers (CNF) with isolated carbonaceous nanoparticles into three-dimensional (3D) conductive frameworks for efficient electron and mass transfer. Herein, a 3D N-doped porous carbon electrocatalyst consisting of carbon nanofibers with grape-like N-doped hollow carbon particles (CNF@NC) has been prepared through a simple electrospinning strategy combined with in-situ growth and carbonization processes. Such a bio-inspired hierarchically organized conductive network largely facilitates both the mass diffusion and electron transfer during the oxygen reduction reactions (ORR). Therefore, the metal-free CNF@NC catalyst demonstrates superior catalytic activity with an absolute four-electron transfer mechanism, strong methanol tolerance and good long-term stability towards ORR in alkaline media.

  16. Tunable superstructure fiber Bragg grating with chirp-distribution modulation based on the effect of external stress.

    Science.gov (United States)

    Huang, Yize; Li, Yi; Zhu, Huiqun; Tong, Guoxiang; Fang, Baoying; Li, Liu; Shen, Yujian; Zheng, Qiuxin; Liang, Qian; Yan, Meng; Wang, Feng; Qin, Yuan; Ding, Jie; Wang, Xiaohua

    2012-09-15

    We report an external stress modulation method for producing a superstructure fiber Bragg grating (FBG) with approximate cascaded resonant cavities composed of different index chirp distributions. The 15 mm uncoated apodized uniform-period FBG is pressed by the vertical stress from the upper 11 pieces of the pattern plate controlled by a piezoelectric ceramic actuator. The piece length is 1 mm, and the interval of the adjacent pieces is 0.4 mm. The reflectivity of the modulated FBG gradually shows six obvious multichannel 75%-85% reflection peaks with the increase of the vertical stress of each pattern-plate piece from 0 to 30 N. The channel spacing is steady at about 10 GHz for a C-band wavelength division multiplexing system.

  17. Diamondlike carbon coating as a galvanic corrosion barrier between dental implant abutments and nickel-chromium superstructures.

    Science.gov (United States)

    Ozkomur, Ahmet; Erbil, Mehmet; Akova, Tolga

    2013-01-01

    The objectives of this study were to evaluate the galvanic corrosion behavior between titanium and nickel-chromium (Ni-Cr) alloy, to investigate the effect of diamondlike carbon (DLC) coating over titanium on galvanic corrosion behavior between titanium and Ni-Cr alloy, and to evaluate the effect of DLC coating over titanium abutments on the fit and integrity of prosthetic assemblies by scanning electron microcopy (SEM). Five Ni-Cr and 10 titanium disks with a diameter of 5 mm and thickness of 3 mm were prepared. DLC coating was applied to five titanium disks. Electrode samples were prepared, and open circuit potential measurements, galvanic current measurements over platinum electrodes, and potentiodynamic polarization tests were carried out. For the SEM evaluation, 20 Ni-Cr alloy and 10 gold alloy superstructures were cast and prepared over 30 abutments. DLC coating was applied to 10 of the abutments. Following the fixation of prosthetic assemblies, the samples were embedded in acrylic resin and cross sectioned longitudinally. Internal fit evaluations were carried out through examination of the SEM images. Titanium showed more noble and electrochemically stable properties than Ni-Cr alloy. DLC coating over the cathode electrode served as an insulating film layer over the surface and prevented galvanic coupling. Results of the SEM evaluations indicated that the DLC-coated and titanium abutments showed no statistically significant difference in fit. Hence, no adverse effects on the adaptation of prosthetic components were found with the application of DLC coating over abutment surfaces. DLC coating might serve as a galvanic corrosion barrier between titanium abutments and Ni-Cr superstructures.

  18. Synthesis, morphology and microstructure of pomegranate-like hematite (α-Fe2O3) superstructure with high coercivity

    International Nuclear Information System (INIS)

    Tadic, Marin; Citakovic, Nada; Panjan, Matjaz; Stanojevic, Boban; Markovic, Dragana; Jovanovic, Đorđe; Spasojevic, Vojislav

    2012-01-01

    Highlights: ► We found superior magnetic properties of the hematite (α-Fe 2 O 3 ). ► TEM and HRTEM images show a pomegranate-like superstructure. ► Magnetic measurements display high coercivity H C = 4350 Oe at the room temperature. - Abstract: We found novel and superior magnetic properties of the hematite (α-Fe 2 O 3 ) that originate from an internal microstructure of particles and strong inter-particle interactions between nanocrystal sub-units. The hematite particles were synthesized by thermal decomposition of iron (III) nitrate without any template or surfactant. The purity, size, crystallinity, morphology, microstructure and magnetic features of the as-prepared particles were characterized by X-ray diffraction (XRD), scanning electron microscope (SEM), energy-dispersive X-ray spectroscopy (EDX), transmission electron microscopy (TEM), high resolution transmission electron microscopy (HRTEM), Raman spectroscopy (RS) and SQUID magnetometry. An XRD study reveals a pure phase of α-Fe 2 O 3 whereas TEM shows α-Fe 2 O 3 spheres with a diameter of about 150 nm. RS also shows high quality and purity of the sample. Moreover, TEM and HRTEM images show a pomegranate-like superstructure and evidence that the spherical particles are composed of individual well-crystallized nanoparticle sub-units (self-assembled nanoparticles) with a size of about 20 nm. Magnetic measurements display hysteretic behavior at the room temperature with remanent magnetization M r = 0.731 emu/g, saturation magnetization M S = 6.83 emu/g and coercivity H C = 4350 Oe, as well as the Morin transition at T M = 261 K. These results and comparison with those in the literature reveal that the sample has extremely high coercivity. The magnetic properties of the sample are discussed in relation to morphology, internal microstructure, surface effects and exchange and dipole–dipole interactions.

  19. Underwater Coatings Testing for INEEL Fuel Basin Applications

    International Nuclear Information System (INIS)

    Julia L. Tripp

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is deactivating several fuel storage basins. Airborne contamination is a concern when the sides of the basins are exposed and allowed to dry during water removal. One way of controlling this airborne contamination is to fix the contamination in place while the pool walls are still submerged. There are many underwater coatings available on the market that are used in marine, naval and other applications. A series of tests were run to determine whether the candidate underwater fixatives are easily applied and adhere well to the substrates (pool wall materials) found in INEEL fuel pools. The four pools considered included (1) Test Area North (TAN-607) with epoxy painted concrete walls; (2) Idaho Nuclear Technology and Engineering Center (INTEC) (CPP-603) with bare concrete walls; (3) Materials Test Reactor (MTR) Canal with stainless steel lined concrete walls; and (4) Power Burst Facility (PBF-620) with stainless steel lined concrete walls on the bottom and epoxy painted carbon steel lined walls on the upper portions. Therefore, the four materials chosen for testing included bare concrete, epoxy painted concrete, epoxy painted carbon steel, and stainless steel. The typical water temperature of the pools varies from 55 F to 80 F dependent on the pool and the season. These tests were done at room temperature

  20. Underwater Coatings Testing for INEEL Fuel Basin Applications

    Energy Technology Data Exchange (ETDEWEB)

    Julia L. Tripp

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is deactivating several fuel storage basins. Airborne contamination is a concern when the sides of the basins are exposed and allowed to dry during water removal. One way of controlling this airborne contamination is to fix the contamination in place while the pool walls are still submerged. There are many underwater coatings available on the market that are used in marine, naval and other applications. A series of tests were run to determine whether the candidate underwater fixatives are easily applied and adhere well to the substrates (pool wall materials) found in INEEL fuel pools. The four pools considered included (1) Test Area North (TAN-607) with epoxy painted concrete walls; (2) Idaho Nuclear Technology and Engineering Center (INTEC) (CPP-603) with bare concrete walls; (3) Materials Test Reactor (MTR) Canal with stainless steel lined concrete walls; and (4) Power Burst Facility (PBF-620) with stainless steel lined concrete walls on the bottom and epoxy painted carbon steel lined walls on the upper portions. Therefore, the four materials chosen for testing included bare concrete, epoxy painted concrete, epoxy painted carbon steel, and stainless steel. The typical water temperature of the pools varies from 55 F to 80 F dependent on the pool and the season. These tests were done at room temperature.

  1. Residual basins

    International Nuclear Information System (INIS)

    D'Elboux, C.V.; Paiva, I.B.

    1980-01-01

    Exploration for uranium carried out over a major portion of the Rio Grande do Sul Shield has revealed a number of small residual basins developed along glacially eroded channels of pre-Permian age. Mineralization of uranium occurs in two distinct sedimentary units. The lower unit consists of rhythmites overlain by a sequence of black shales, siltstones and coal seams, while the upper one is dominated by sandstones of probable fluvial origin. (Author) [pt

  2. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

    Howell, J.P.; Burke, S.D.

    1996-01-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy

  3. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    Energy Technology Data Exchange (ETDEWEB)

    Howell, J.P.; Burke, S.D.

    1996-02-20

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.

  4. Hierarchical super-structure identified by polarized light microscopy, electron microscopy and nanoindentation: Implications for the limits of biological control over the growth mode of abalone sea shells

    Directory of Open Access Journals (Sweden)

    Schneider Andreas S

    2012-09-01

    Full Text Available Abstract Background Mollusc shells are commonly investigated using high-resolution imaging techniques based on cryo-fixation. Less detailed information is available regarding the light-optical properties. Sea shells of Haliotis pulcherina were embedded for polishing in defined orientations in order to investigate the interface between prismatic calcite and nacreous aragonite by standard materialographic methods. A polished thin section of the interface was prepared with a defined thickness of 60 μm for quantitative birefringence analysis using polarized light and LC-PolScope microscopy. Scanning electron microscopy images were obtained for comparison. In order to study structural-mechanical relationships, nanoindentation experiments were performed. Results Incident light microscopy revealed a super-structure in semi-transparent regions of the polished cross-section under a defined angle. This super-structure is not visible in transmitted birefringence analysis due to the blurred polarization of small nacre platelets and numerous organic interfaces. The relative orientation and homogeneity of calcite prisms was directly identified, some of them with their optical axes exactly normal to the imaging plane. Co-oriented "prism colonies" were identified by polarized light analyses. The nacreous super-structure was also visualized by secondary electron imaging under defined angles. The domains of the super-structure were interpreted to consist of crystallographically aligned platelet stacks. Nanoindentation experiments showed that mechanical properties changed with the same periodicity as the domain size. Conclusions In this study, we have demonstrated that insights into the growth mechanisms of nacre can be obtained by conventional light-optical methods. For example, we observed super-structures formed by co-oriented nacre platelets as previously identified using X-ray Photo-electron Emission Microscopy (X-PEEM [Gilbert et al., Journal of the

  5. Nb2OsB2, with a new twofold superstructure of the U3Si2 type: Synthesis, crystal chemistry and chemical bonding

    International Nuclear Information System (INIS)

    Mbarki, Mohammed; Touzani, Rachid St.; Fokwa, Boniface P.T.

    2013-01-01

    The new ternary metal-rich boride, Nb 2 OsB 2 , was synthesized by arc-melting the elements in a water-cooled copper crucible under an argon atmosphere. The compound was characterized from single-crystal X-ray data and EDX measurements. It crystallizes as a new superstructure (space group P4/mnc, no. 128) of the tetragonal U 3 Si 2 -structure type with lattice parameters a=5.922(1) Å and c=6.879(2) Å. All of the B atoms are involved in B 2 dumbbells with B–B distances of 1.89(4) Å. Structure relaxation using VASP (Vienna ab intio Simulation Package) has confirmed the space group and the lattice parameters. According to electronic structure calculations (TB–LMTO–ASA), the homoatomic B–B interactions are optimized and very strong, but relatively strong heteroatomic Os–B, Nb–B and Nb–Os bonds are also found: These interactions, which together build a three-dimensional network, are mainly responsible for the structural stability of this new phase. The density of state at the Fermi level predicts metallic behavior, as expected, from this metal-rich boride. - Graphical abstract: Nb 2 OsB 2 is, to the best of our knowledge, the first fully characterized phase in the ternary Nb–Os–B system. It crystallizes (space group P4/mnc, 128) with a new twofold superstructure of the U 3 Si 2 structure type (space group P4/mbm, 127), and is therefore the first boride in this structure family crystallizing with a superstructure of the U 3 Si 2 structure type. We show that the distortions leading to this superstructure occurs mainly in the Nb-layer, which tries to accommodate the large osmium atoms. The consequence of this puckering is the building osmium dumbbells instead of chains along [001]. - Highlights: • First compound in the Nb–Os–B system. • New twofold superstructure of U 3 Si 2 structure type. • Puckering of Nb-layer responsible for superstructure occurrence. • Chemical bonding studied by density functional theory

  6. The Origin of the Superstructure in Bi2Sr2CaCu2O8+dgr as Revealed by Scanning Tunneling Microscopy.

    Science.gov (United States)

    Kirk, M D; Nogami, J; Baski, A A; Mitzi, D B; Kapitulnik, A; Geballe, T H; Quate, C F

    1988-12-23

    Real-space images with atomic resolution of the BiO plane of Bi(2)Sr(2)CaCu(2)O(8+delta) were obtained with a scanning tunneling microscope. Single-crystal samples were cleaved and imaged under ultrahigh vacuum conditions at room temperature. The images clearly show the one-dimensional incommensurate superstructure along the b-axis that is common to this phase. High-resolution images show the position of the Bi atoms, revealing the structural nature of the superlattice. A missing row of Bi atoms occurs either every nine or ten atomic sites in both (110) directions, accounting for the measured incommensurate periodicity of the superstructure. A model is proposed that includes missing rows of atoms, as well as displacements of the atomic positions along both the a- and c-axis directions.

  7. The origin of the superstructure in Bi2Sr2CaCu2O(8+delta) as revealed by scanning tunneling microscopy

    Science.gov (United States)

    Kirk, M. D.; Nogami, J.; Baski, A. A.; Mitzi, D. B.; Kapitulnik, A.

    1988-12-01

    Real-space images with atomic resolution of the BiO plane of Bi2Sr2CaCu2O(8+delta) were obtained with a scanning tunneling microscope. Single-crystal samples were cleaved and imaged under ultrahigh vacuum conditions at room temperature. The images clearly show the one-dimensional incommensurate superstructure along the b-axis that is common to this phase. High-resolution images show the position of the Bi atoms, revelaing the structural nature of the superlattice. A missing row of Bi atoms occurs either every nine or ten atomic sites in both 110-line directions, accounting for the measured incommensurate periodicity of the superstructure. A model is proposed that includes missing rows of atoms, as well as displacements of the atomic positions along both the a- and c-axis directions.

  8. Synthesis and superstructure of La sub(2/3) (Mg sub(1/2)W sub(1/2))O/sub 3/

    Energy Technology Data Exchange (ETDEWEB)

    Torii, Y [Government Industrial Research Inst., Nagoya (Japan)

    1979-10-01

    A new perovskite La sub(2/3)(Mg sub(1/2)W sub(1/2))O/sub 3/ having an orthorhombic multiple-cell was synthesized. The lattice constants were a = 7.8157(5) A, b = 7.8344(6) A and c = 2 x 7.9067(6) A. The superstructure was found to be due to a NaCl-type ordering of B ions as well as an ordering of A-site vacancies.

  9. Giant reversible magnetocaloric effect in flower-like β-Co(OH){sub 2} hierarchical superstructures self-assembled by nanosheets

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xianguo; Feng, Chao; Xiao, Feng; Jin, Chuangui; Xia, Ailin, E-mail: liuxianguohugh@gmail.com, E-mail: eeswor@polyu.edu.hk [Anhui Key Laboratory of Metal Materials and Processing, School of Materials Science and Engineering, Anhui University of Technology, Ma' anshan, PR (China); Or, Siu Wing [Department of Electrical Engineering, The Hong Kong Polytechnic University, Hung Hom, Kowloon (Hong Kong); Sun, Yuping [Center for Engineering Practice and Innovation Education, Anhui University of Technology, Ma' anshan, PR (China)

    2014-01-15

    A facile hydrothermal strategy is proposed to synthesize flower-like β- Co(OH){sub 2} hierarchical microspherical superstructures with a diameter of 0.5-1.5 µm, which are self-assembled by β - Co(OH){sub 2} nano sheets with the average thickness ranging between 20 and 40 nm. The magnetocaloric effect associated with magnetic phase transitions in Co(OH){sub 2} superstructures has been investigated. A sign change in the magnetocaloric effect is induced by a magnetic field, which is related to a filed-induced transition from the antiferromagnetic to the ferromagnetic state below the Néel temperature. The large reversible magnetic-entropy change –ΔS{sub m} (13.4 J/kg K at 15 K for a field change of 5 T) indicates that flower-like Co(OH){sub 2} superstructures is a potential candidate for application in magnetic refrigeration in the low-temperature range. (author)

  10. Topotactic Epitaxy of SrTiO3 Mesocrystal Superstructures with Anisotropic Construction for Efficient Overall Water Splitting.

    Science.gov (United States)

    Zhang, Peng; Ochi, Tomoya; Fujitsuka, Mamoru; Kobori, Yasuhiro; Majima, Tetsuro; Tachikawa, Takashi

    2017-05-02

    The higher-order structures of semiconductor-based photocatalysts play crucial roles in their physicochemical properties for efficient light-to-energy conversion. A novel perovskite SrTiO 3 mesocrystal superstructure with well-defined orientation of assembled cubic nanocrystals was synthesized by topotactic epitaxy from TiO 2 mesocrystals through a facile hydrothermal treatment. The SrTiO 3 mesocrystal exhibits three times the efficiency for the hydrogen evolution of conventional disordered systems in alkaline aqueous solution. It also exhibits a high quantum yield of 6.7 % at 360 nm in overall water splitting and even good durability up to 1 day. Temporal and spatial spectroscopic observations revealed that the synergy of the efficient electron flow along the internal nanocube network and efficient collection at the larger external cubes produces remarkably long-lived charges for enhanced photocatalysis. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Yb{sub 6}Ir{sub 5}Ga{sub 7} - a MgZn{sub 2} superstructure

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, Stefan; Poettgen, Rainer [Institut fuer Anorganische und Analytische Chemie, Universitaet Muenster, Corrensstrasse 30, 48149, Muenster (Germany)

    2017-02-15

    The gallide Yb{sub 6}Ir{sub 5}Ga{sub 7} was synthesized by high-frequency melting of the elements in a sealed niobium ampoule. The structure was refined from single-crystal X-ray diffractometer data: Nb{sub 6.4}Ir{sub 4}Al{sub 7.6} type, P6{sub 3}/mcm, a = 930.4(1), c = 843.0(1) pm, wR{sub 2} = 0.0597, 379 F{sup 2} values and 22 variables. Yb{sub 6}Ir{sub 5}Ga{sub 7} adopts a superstructure of the MgZn{sub 2} Laves phase by a complete ordering of the iridium and gallium atoms on the zinc substructure, i.e. the network consists of ordered and condensed Ir{sub 3}Ga and IrGa{sub 3} tetrahedra with Ir-Ga distances ranging from 260 to 265 pm. The crystal chemical details and the underlying group-subgroup scheme are discussed. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Investigating the construction of open-quotes pyramidclose quotes super-structures to dispose of radioactive and hazardous waste

    International Nuclear Information System (INIS)

    Miller, D.J.

    1994-01-01

    The technical challenges involved in disposing of radioactive and hazardous waste using systems that rely on open-quotes natural barriersclose quotes have been underestimated. Technology has advanced dramatically in the areas of materials, science, and engineering. As a result, traditional approaches to waste disposal must be rethought, focusing instead on ways to apply new technology breakthroughs to waste disposal problems. This paper will discuss problems currently faced by waste disposal systems that rely on natural barriers for containment, propose a general design and approach for constructing pyramid super-structures to dispose of (and store) waste, and present the benefits of such a system. The originality of this paper is that it proposes the construction of fully retrievable waste disposal systems that capitalize on the unique geometric properties of a pyramid and rely on engineered barriers and preventive measurements for containment, rather than natural barriers. In addition, this paper offers a new perspective on waste disposal issues confronting many countries. The desired effect is that by challenging conventional thought, new ideas could be developed to help solve existing problems. This paper is of specific interest to: (1) Policy makers, decisionmakers, and managers because the paper discusses the root causes of problems facing waste disposal and possible solutions. (2) Regulators because the paper discusses reasons for relying more on engineered barriers and preventive measurements to achieve confidence and reliability in containment. (3) Design engineers the paper offers new concepts for engineered barriers. 21 refs., 2 figs., 2 tabs

  13. Proposal for the award of an industrial services contract for civil engineering superstructure and internal construction building work

    CERN Document Server

    2001-01-01

    This document concerns the award of an Industrial Services contract for civil engineering superstructure and internal construction building work. Following a market survey carried out among 112 firms in fifteen Member States, a call for tenders (IT-2546/ST/Rev.) was sent on 12 April 2001 to eleven consortia, four consisting of four firms, three consisting of three firms and four consisting of two firms, in seven Member States. By the closing date, CERN had received tenders from eight consortia in six Member States. The Finance Committee is invited to agree to the negotiation of a contract with the consortium SPIE CITRA SUD-EST(FR)-ANTIRUST TECHNOLOGY (GR), the lowest bidder, for an initial period of three years from 1 January 2002 for an amount not exceeding 15 375 000 Swiss francs, not subject to revision. The contract will include options for two one-year extensions beyond the initial three-year period. The consortium has indicated the following distribution by country of the contract value covered by this ...

  14. Ga induced superstructures as templates for lattice matched hetroepitaxial growth of GaN on Si(111) substrate

    International Nuclear Information System (INIS)

    Kumar, Praveen; Kuyyalil, Jithesh; Shivaprasad, S. M.

    2010-01-01

    High quality GaN is grown by plasma assisted molecular beam epitaxy on Ga induced superstructural phases of Si(111)7x7. Three stable surface phases induced by Ga adsorption, viz., (1x1), (6.3x6.3), and (√3x√3)R30 deg., are employed as templates to grow epitaxial (0001) GaN thin films. GaN grown on Si(√3x√3)R30 deg. -Ga is found to be highly crystalline with intense (0002) x-ray diffraction and photoluminescence peaks with low full width at half maximum, low surface roughness, and stoichiometric surface composition. The high quality of these GaN films formed at a low temperature of 400 deg. C is explained by the integral (x2) lattice matching between the unit cell of GaN and the (√3x√3) phase. The experiments demonstrate a plausible approach of adsorbate induced surface modifications as templates for III-V hetroepitaxy on Si surfaces.

  15. Characterization of basin concrete in support of structural integrity demonstration for extended storage

    International Nuclear Information System (INIS)

    Duncan, A.

    2014-01-01

    Concrete core samples from C basin were characterized through material testing and analysis to verify the design inputs for structural analysis of the L Basin and to evaluate the type and extent of changes in the material condition of the concrete under extended service for fuel storage. To avoid the impact on operations, core samples were not collected from L area, but rather, several concrete core samples were taken from the C Basin prior to its closure. C basin was selected due to its similar environmental exposure and service history compared to L Basin. The microstructure and chemical composition of the concrete exposed to the water was profiled from the water surface into the wall to evaluate the impact and extent of exposure. No significant leaching of concrete components was observed. Ingress of carbonation or deleterious species was determined to be insignificant. No evidence of alkali-silica reactions (ASR) was observed. Ettringite was observed to form throughout the structure (in air voids or pores); however, the sulfur content was measured to be consistent with the initial concrete that was used to construct the facility. Similar ettringite trends were observed in the interior segments of the core samples. The compressive strength of the concrete at the mid-wall of the basin was measured, and similar microstructural analysis was conducted on these materials post compression testing. The microstructure was determined to be similar to near-surface segments of the core samples. The average strength was 4148 psi, which is well-above the design strength of 2500 psi. The analyses showed that phase alterations and minor cracking in a microstructure did not affect the design specification for the concrete

  16. Unreviewed safety question evaluation of 100K East and 100K West in-basin fuel characterization program activities

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1995-01-01

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation of the 105K East (KE) and 105K West (KW) in-basin activities associated with the fuel characterization program as described in the characterization shipping plan. The significant activities that are common to both 105 KE and 105 KW basins are the movement of canisters from their main basin storage locations (or potentially from the 105 KE Tech View Pit if a dump table is available) to the south loadout pit transfer channel, hydrogen generation testing in the single element fuel container, loading the single element fuel container into the shipping cask, loading of the shipping cask onto a flat-bed trailer, return of the test fuel elements or element pieces from the 327 facility, placement of the fuel elements back into Mark 2 canisters, and placement of the canisters in the main storage basin. Decapping of canisters in the south loadout pit transfer channel and re-encapsulation of canisters are activities specific to the 105 KW basin. The scope of this safety evaluation includes only those characterization fuel shipment activities performed in the 105 KE and 105 KW fuel storage basin structures up to installation of the overpack. The packaging safety evaluation report governs the shipment of the fuel elements. The K Basins Plant Review Committee has determined that the in-basin activities associated with the fuel characterization program fuel shipments are bounded by the current safety envelop and do not constitute an unreviewed safety question. This determination is documented on Attachment 1

  17. A multi-period superstructure optimisation model for the optimal planning of China's power sector considering carbon dioxide mitigation

    International Nuclear Information System (INIS)

    Zhang Dongjie; Ma Linwei; Liu Pei; Zhang Lili; Li Zheng

    2012-01-01

    Power sector is the largest CO 2 emitter in China. To mitigate CO 2 emissions for the power sector is a tough task, which requires implementation of targeted carbon mitigation policies. There might be multiple forms for carbon mitigation policies and it is still unclear which one is the best for China. Applying a superstructure optimisation model for optimal planning of China's power sector built by the authors previously, which was based on real-life plants composition data of China's power sector in 2009, and could incorporate all possible actions of the power sector, including plants construction, decommission, and application of carbon capture and sequestration (CCS) on coal-fuelled plants, the implementation effects of three carbon mitigation policies were studied quantitatively, achieving a conclusion that the so-called “Surplus-Punishment and Deficit-Award” carbon tax policy is the best from the viewpoint of increasing CO 2 reduction effect and also reducing the accumulated total cost. Based on this conclusion, the corresponding relationships between CO 2 reduction objectives (including the accumulated total emissions reduction by the objective year and the annual emissions reduction in the objective year) were presented in detail. This work provides both directional and quantitative suggestions for China to make carbon mitigation policies in the future. - Highlights: ► We study the best form of carbon mitigation policy for China's power sector. ► We gain quantitative relationship between CO 2 reduction goal and carbon tax policy. ► The “Surplus-Punishment and Deficit-Award” carbon tax policy is the best. ► Nuclear and renewable power and CCS can help greatly reduce CO 2 emissions of the power sector. ► Longer objective period is preferred from the viewpoint of policy making.

  18. Subproject plan for demonstration of 3M technology for treatment of N Basin water

    International Nuclear Information System (INIS)

    Plastino, J.C.

    1996-02-01

    A dissolved radionuclides removal demonstration is being conducted at the 105-N Basin as part of the 100-N Area Projects' policy of aggressively integrating innovative technologies to achieve more cost effective, faster, and/or safer deactivation operations. This subproject plan demonstrates new technology (marketed by the 3M trademark Company) that absorbs specific ions from water. The demonstration will take place at the spent fuel basin at the N Reactor facility. The 105-N Basin contains 1 million gal of water consisting of approximately 32 Ci of dissolved 90 Sr at a concentration of 8.4 uCi/L and 7.3 Ci of dissolved 137 Cs at a concentration of 1.92 uCi/L. The Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement [Ecology et al. 1990]) Milestone M-16-01E-T2 requires the initiation of pretreatment and removal of all N Reactor fuel storage basin waters by September 30, 1996, pursuant to the N Reactor Deactivation Program Plan (WHC 1993). 105-N Basin dewatering is on the critical path for overall deactivation of N Reactor by March 1997. The 105-N Basin Deactivation Program Plan (BHI 1995) includes removing debris, hardware, algae and sediment from the basin, followed by pretreatment (filtration) and removal of the 1005-N Basin water. Final water removal is currently scheduled for September 30, 1996. The recommended method of the 105-N Basin water is the treatment of the water at the Effluent Treatment Facility (ETF) in the 200 East Area. The demonstration of the 3M technology could be a feasible treatment alternative to the ETF if the ETF is not available to meet the project schedule or if additional pretreatment is needed to reduce the inventory of radioactive species to be handled at the ETF. Demonstration of this technology could be of value for other fuel basins at the Hanford Site and possibly other US Department of Energy (DOE) sites and non- DOE nuclear power plants

  19. Final environmental impact statement for the construction and operation of an independent spent fuel storage installation to store the Three Mile Island Unit 2 spent fuel at the Idaho National Engineering and Environmental Laboratory. Docket Number 72-20

    International Nuclear Information System (INIS)

    1998-03-01

    This Final Environmental Impact Statement (FEIS) contains an assessment of the potential environmental impacts of the construction and operation of an Independent Spent Fuel Storage Installation (ISFSI) for the Three Mile Island Unit 2 (TMI-2) fuel debris at the Idaho National Engineering and Environmental laboratory (INEEL). US Department of Energy-Idaho Operations Office (DOE-ID) is proposing to design, construct, and operate at the Idaho Chemical Processing Plant (ICPP). The TMI-2 fuel debris would be removed from wet storage, transported to the ISFSI, and placed in storage modules on a concrete basemat. As part of its overall spent nuclear fuel (SNF) management program, the US DOE has prepared a final programmatic environmental impact statement (EIS) that provides an overview of the spent fuel management proposed for INEEL, including the construction and operation of the TMI-2 ISFSI. In addition, DOE-ID has prepared an environmental assessment (EA) to describe the environmental impacts associated with the stabilization of the storage pool and the construction/operation of the ISFSI at the ICPP. As provided in NRC's NEPA procedures, a FEIS of another Federal agency may be adopted in whole or in part in accordance with the procedures outlined in 40 CFR 1506.3 of the regulations of the Council on Environmental Quality (CEQ). Under 40 CFR 1506.3(b), if the actions covered by the original EIS and the proposed action are substantially the same, the agency adopting another agency's statement is not required to recirculate it except as a final statement. The NRC has determined that its proposed action is substantially the same as actions considered in DOE's environmental documents referenced above and, therefore, has elected to adopt the DOE documents as the NRC FEIS

  20. Demonstration of radionuclides removal at the 105-N basin using the 3M system

    International Nuclear Information System (INIS)

    Hobart, S.A.; Hyman, M.

    1996-03-01

    A field demonstration of the Minnesota Mining and Manufacturing Company (3M) innovative Cs- and Sr-removal technology was undertaken to support deactivation of the Hanford Site's N Reactor. The field demonstration and laboratory studies performed by Pacific Northwest National Laboratory are documented in this report. The deactivation of N Reactor includes an aggressive schedule for removing water from the reactor's original fuel storage pool, known as N Basin. The plan for treating N Basin water involves particulate filtration, followed by further treatment at the Hanford Site's Effluent Treatment Facility (ETF) before the water is discharged to a permitted soil column. Prudence calls for developing a backup treatment plan in the event that ETF is not available to support the N Basin critical path schedule. The 3M technique uses membrane filters that are impregnated with chemical species to remove specific ions from water by ion exchange or selective reaction. Sodium titanate is used to remove Sr by ion exchange, and potassium-cobalt hexacyanoferrate (COHEX) is used to remove Cs by formation of cesium-cobalt ferrocyanide. As a result of this field test, the following recommendations are made: 3M technology should not be considered for removing Sr from N Basin water, although the technology merits consideration for Sr removal in waters that have relatively low Cs content; application of 3M technology by recirculation of basin water through 3M adsorption cartridges for removal of Cs-137 should be considered since it efficiently removes Cs and may be cost effective; additional pilot-scale tests should be performed to determine the capacity of COHEX for Cs-137 removal, if full-scale application is desired