DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

International Nuclear Information System (INIS)

J. Bisset

2005-01-01

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

Concrete storage cask for interim storage of spent nuclear fuel

International Nuclear Information System (INIS)

Nabemoto, Toyonobu; Fujiwara, Hiroaki; Kobayashi, Shunji; Shionaga, Ryosuke

2004-01-01

Experiments and analytical evaluation of the fabrication, non-destructive inspection and structural integrity of reinforced concrete body for storage casks were carried out to demonstrate the concrete storage cask for spent fuel generated from nuclear power plants. Analytical survey on the type of concrete material and fabrication method of the storage cask was performed and the most suitable fabrication method for the concrete body was identified to reduce concrete cracking. The structural integrity of the concrete body of the storage cask under load conditions during storage was confirmed and the long term integrity of concrete body against degradation dependent on environmental factors was evaluated. (author)

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...

Developing new transportable storage casks for interim dry storage

Energy Technology Data Exchange (ETDEWEB)

Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R. [Hitachi Zosen Diesel and Engineering Co., Ltd., Tokyo (Japan)

2004-07-01

Transportable storage metal casks are to be consistently used during transport and storage for AFR interim dry storage facilities planning in Japan. The casks are required to comply with the technical standards of regulations for both transport (hereinafter called ''transport regulation'') and storage (hereafter called ''storage regulation'') to maintain safety functions (heat transfer, containment, shielding and sub-critical control). In addition to these requirements, it is not planned in normal state to change the seal materials during storage at the storage facility, therefore it is requested to use same seal materials when the casks are transported after storage period. The dry transportable storage metal casks that satisfy the requirements have been developed to meet the needs of the dry storage facilities. The basic policy of this development is to utilize proven technology achieved from our design and fabrication experience, to carry out necessary verification for new designs and to realize a safe and rational design with higher capacity and efficient fabrication.

CASTOR registered HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

International Nuclear Information System (INIS)

Vossnacke, A.; Klein, K.; Kuehne, B.

2004-01-01

Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR registered HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR registered HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR registered HAW28M, meeting these future requirements. The CASTOR registered HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR registered HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be considered. For the CASTOR

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...

Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

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...

Low-cost concepts for dry transfer of spent fuel and waste between storage and transportation casks

International Nuclear Information System (INIS)

Schneider, K.L.

1984-01-01

The federal government may provide interim storage for spent fuel from commercial nuclear power reactors that have used up their available storage capacity. One of the leading candidate concepts for this interim storage is to place spent fuel in large metal shielding casks. The Federal Interim Storage (FIS) site may not have the capability to transfer spent fuel from transportation casks to storage casks and vice versa. Thus, there may be an incentive to construct a relatively inexpensive but reliable intercask transfer system for use at an FIS site. This report documents the results of a preliminary study of preconceptual design and analysis of four intercask transfer concepts. The four concepts are: a large shielded cylindrical turntable that contains an integral fuel handling machine (turntable concept); a shielded fuel handling machine under which shipping and storage casks are moved horizontally (shuttle concept); a small hot cell containing equipment for transferring fuel betwee shipping and storage casks (that enter and leave the cell on carts) in a bifurcated trench (trench concept) and a large hot cell, shielded by an earthen berm, that houses equipment for handling fuel between casks that enter and leave the cell on a single cart (igloo concept). Information derived for each of the concepts is operating, capital and relocation costs; implementation and relocation time requirements; and overall characteristics

CASTOR {sup registered} HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

Energy Technology Data Exchange (ETDEWEB)

Vossnacke, A.; Klein, K.; Kuehne, B. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany)

2004-07-01

Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR {sup registered} HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR {sup registered} HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR {sup registered} HAW28M, meeting these future requirements. The CASTOR {sup registered} HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR {sup registered} HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2011-01-01

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

Operational and safety aspects of vitrified waste casks

International Nuclear Information System (INIS)

Kirchner, B.

1993-01-01

For the time being two technical solutions have been developed for the interim storage: 1) one is based on forced air cooled pits set out in a concrete structure, as presently provided close to the Vitrification Facilities on reprocessing sites; 2) the other one is based on transportable storage casks standing vertically onto a storage pad, following principles similar to those already experienced with spent fuel storage casks. Considering these two solutions for interim storage, TRANSNUCLEAIRE has developed two main types of transportable casks for vitrified HAW; one is a routine transport cask; the other one is a transportable storage cask. Both are covered by the generic name TN28V and have already been described in previous papers. This paper deals with the safety and operation aspects of the casks under both transport and storage conditions. (J.P.N.)

Dry storage systems using casks for long term storage in an AFR and repository

International Nuclear Information System (INIS)

Einfeld, K.; Popp, F.W.

1986-01-01

In conclusion it can be stated that two basic routes with respect to spent fuel storage casks are feasible. One is the Multiple Transport Cask, which with certain modifications can be upgraded to meet the criteria for intermediate storage. Its status is characterized by the licensing of several types of Castor Casks for an intermediate storage period of 30 years in the AFR Storage Facility of DWK at Gorleben in the FRG. The other one is the Final Disposal (Repository) Cask, which can be made suitable for long term storage before a final decision with respect to a repository application is taken. The licensing procedure for a Pilot Conditioning Facility with the Pollux Cask System as reference case will be initiated by DWK in the near future. Under the assumption that in addition to the present Multiple Transport/Storage Casks a license for a Final disposal Cask with respect to long term storage is available, the relative merits of different cask storage systems would have to be evaluated

Monitored Retrievable Storage conceptual system study: metal storage casks

International Nuclear Information System (INIS)

Unterzuber, R.; Cross, T.E.; Krasicki, B.R.

1983-08-01

A description of the metal cask storage facility concept is presented with the operations required to handle the spent fuel or high-level wastes and transuranic wastes. A generic Receiving and Handling Facility, provided by PNL, has been used for this study. Modifications to the storage delivery side of the handling facility, necessary to couple the Receiving and Handling Facility with the storage facility, are described. The equipment and support facilities needed for the storage facility are also described. Two separate storage facilities are presented herein: one for all spent fuel storage, and one for storage of high-level waste (HLW) and transuranic waste (TRU). Each facility is described for the capacities and rates defined by PNL in the Concept Technical Performance Criteria and Base Assumptions (see Table 1.3-1). Estimates of costs and time-distributions of expenditures have been developed to construct, operate, and decommission the conceptual MRS facilities in mid-1983 dollars, for the base cases given using the cost categories and percentages provided by PNL. Cost estimates and time-distributions of expenditures have also been developed to expand the facility throughput rate from 1800 MTU to 3000 MTU, and to expand the facility storage capacity from 15,000 MTU to 72,00 MTU. The life cycle cost of the facility for the bounding cases of all spent fuel and all HLW and TRU, using the time-distributions of costs developed above and assuming a two percent per year discount rate, are also presented. 3 references, 16 figures, 18 tables

An optimized cask technology for conditioning, transportation and long term interim storage of 'End of Life' nuclear waste

International Nuclear Information System (INIS)

Lefort-Mary, Florence; Clement, Gilles; Lamouroux, Christine; Dumont, Bruno

2016-01-01

When preparing for the decommissioning of a nuclear facility, during its 'end of life' management and while performing the actual dismantling operations, one has to consider a large diversity of nuclear waste in term of types, volumes and activities. Customers are frequently faced with the obligation to undertake multiple and costly waste management operations including handling, reconditioning or re-transferring from one package to another, for example when moving from on-site storage to transportation. To address this issue, a new - highly flexible - cask system named TN R MW is being developed. This cask has a total weight of 10 T and is compliant with the 2012 IAEA regulations. It is developed on a flexible concept basis, adaptable to the various nuclear needs, including: from IP2 to B(U) / B(U)F; on-site/ international transportation; long term interim storage. Licensing and manufacturing of number of items of this TN R MW family is underway. (authors)

COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

International Nuclear Information System (INIS)

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

2007-01-01

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

COMPILATION OF DISPOSABLE SOLID WASTE CASK EVALUATIONS

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-21

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

Standard review plan for dry cask storage systems. Final report

International Nuclear Information System (INIS)

1997-01-01

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

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

International Nuclear Information System (INIS)

Kenneally, R.; Kessler, J.

2001-01-01

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

Source storage and transfer cask: Users Guide

International Nuclear Information System (INIS)

Eccleston, G.W.; Speir, L.G.; Garcia, D.C.

1985-04-01

The storage and shield cask for the dual californium source is designed to shield and transport up to 3.7 mg (2 Ci) of 252 Cf. the cask meets Department of Transportation (DOT) license requirements for Type A materials (DOT-7A). The cask is designed to transfer sources to and from the Flourinel and Fuel Storage (FAST) facility delayed-neutron interrogator. Californium sources placed in the cask must be encapsulated in the SR-CF-100 package and attached to Teleflex cables. The cask contains two source locations. Each location contains a gear box that allows a Teleflex cable to be remotely moved by a hand crank into and out of the cask. This transfer procedure permits sources to be easily removed and inserted into the delayed-neutron interrogator and reduces personnel radiation exposure during transfer. The radiation dose rate with the maximum allowable quantity of californium (3.7 mg) in the cask is 30 mR/h at the surface and less than 2 mR/h 1 m from the cask surface. This manual contains information about the cask, californium sources, describes the method to ship the cask, and how to insert and remove sources from the cask. 28 figs

Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

International Nuclear Information System (INIS)

Shimizu, Masashi; Hayashi, Makoto; Kashiwakura, Jun

2003-01-01

Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

Storage container for radioactive wastes

International Nuclear Information System (INIS)

Catalayoud, L.; Gerard, M.

1990-01-01

Tightness, shock resistance and corrosion resistance of containers for storage of radioactive wastes it obtained by complete fabrication with concrete reinforced with metal fibers. This material is used for molding the cask, the cover and the joint connecting both parts. Dovetail grooves are provided on the cask and the cover for the closure [fr

On-site concrete cask storage system for spent nuclear fuel

International Nuclear Information System (INIS)

Craig, P.A.; Haelsig, R.T.; Kent, J.D.; Schmoker, D.S.

1989-01-01

A method is described of storing spent nuclear fuel assemblies including the steps of: transferring the fuel assemblies from a spent-fuel pool to a moveable concrete storage cask located outside the spent-fuel pool; maintaining a barrier between the fuel and the concrete in the cask to prevent contamination of the concrete by the fuel; maintaining the concrete storage cask containing the spent-fuel on site at the reactor complex for some predetermined period; transferring the fuel assemblies from the concrete storage cask to a shipping container; and, recycling the concrete storage cask

Capabilities for processing shipping casks at spent fuel storage facilities

International Nuclear Information System (INIS)

Baker, W.H.; Arnett, L.M.

1978-01-01

Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

Seismic stability of unanchored spent nuclear fuel storage casks

International Nuclear Information System (INIS)

Ofoegbu, G. I.; Gute, G. D.; Chowdhury, A. H.

2003-01-01

Dynamic soil-structure interaction analyses were performed to examine the effects of a potential earthquake on the stability of unanchored cylindrical spent nuclear fuel casks for an above-ground storage installation. The casks would be placed on a cluster of reinforced concrete pads founded on a deep sequence of clays and silts underlain by sandstones. The analyses focused on evaluating the geometric stability of the casks during an earthquake with respect to a design concept that a cask should not tip over, slide off the storage pad, or collide with another cask. The analyses were performed using LS-DYNA with a three-dimensional explicit finite element model representing the site soil and a fully loaded storage pad. Three statistically independent acceleration time histories were applied simultaneously at the base of the model to generate a free-field ground motion time history representing the design-basis earthquake. Sensitivity studies were performed to examine the effects of the interface conditions between the storage pad and the surrounding soil, and between the base of the storage casks and the top surface of the pad. The results indicate that ground motion from the design-basis earthquake would not cause any cask to tip over, slide off the pad, or collide with another cask. The contact conditions at the cask-to-pad and pad-to-soil interfaces have a strong effect on potential cask motions during an earthquake. If the cask-base friction coefficient is small, the casks may slide, but would not experience any significant rocking. If the cask-base friction is large enough to permit a significant transfer of earthquake lateral motions across the cask-to-pad interface, a design with bonded pad-to-soil interfaces would produce larger cask motions than a design with frictional pad-to-soil interfaces. Furthermore, a cask strage design in which the cask motions are essentially isolated from the motions of the pad-soil system, which can be accomplished if the cask

Concrete spent fuel storage casks dose rates

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Trontl, K.

1998-01-01

Our intention was to model a series of concrete storage casks based on TranStor system storage cask VSC-24, and calculate the dose rates at the surface of the casks as a function of extended burnup and a prolonged cooling time. All of the modeled casks have been filled with the original multi-assembly sealed basket. The thickness of the concrete shield has been varied. A series of dose rate calculations for different burnup and cooling time values have been performed. The results of the calculations show rather conservative original design of the VSC-24 system, considering only the dose rate values, and appropriate design considering heat rejection.(author)

Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

International Nuclear Information System (INIS)

Creer, J.M.; McKinnon, M.A.; Collantes, C.E.

1990-01-01

To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

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...

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...

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

Science.gov (United States)

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

2018-04-01

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

B cell remote-handled waste shipment cask alternatives study

International Nuclear Information System (INIS)

RIDDELLE, J.G.

1999-01-01

The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

Dry cask storage: a Vepco/DOE/EPRI cooperative demonstration program

International Nuclear Information System (INIS)

Smith, M.L.

1984-01-01

In response to a Department of Energy (DOE) Solicitation for Cooperative Agreement Proposal, Virginia Electric and Power Company (Vepco) proposed to participate in a spent fuel storage demonstration program utilizing the dry cask storage technology. This proposed program includes dry cask storage at Vepco's Surry Nuclear Power Station and research and development activities at a DOE site in support of the licensed program at Surry. Phase I of Vepco's two-phase program involves a demonstration of the licensed dry cask storage of spent fuel in an inert atmosphere at the Surry Power Station site. Phase II of Vepco's proposed program will involve the demonstration of storing unconsolidated and consolidated spent fuel in dry casks filled only with air. This phase of the program will involve DOE site testing similar to Phase I and is expected to require an additional (fourth) cask to demonstrate storage of unconsolidated spent fuel in air-filled casks

Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

The state of the Primary Degradation Factors and Models of Concrete Cask in Spent Fuel Dry Storage System

International Nuclear Information System (INIS)

Kim, J. S.; Lee, K. S.; Choi, J. W.; Kwon, S.

2010-01-01

In South Korea, a total of twenty nuclear reactors are in operation; the cumulative amount of spent fuel is estimated to be 10,490 MTU in 2009. The full capacity of the waste storage is expected to be saturated in around 2016. However, a national strategy for spent fuel management has not yet been set down and high level waste (HLW) such as spent fuel will have to be stored at-reactor (AR) by re-racking. Recently an worldwide interest on the dry storage has increased especially around U.S. With a perspective of the material of the spent fuel dry storage cask, the system can be divided into two types of metal and concrete casks. The concrete type cask is a very attractive option because of the cost competitiveness of concrete material and its relatively long-term durability. Although the type of metal cask is chosen, the use of cementitious material is inevitable at least for the cask foundation and the facilities for the protection of dry storage structures. Upon being placed, the performance of concrete begins to deteriorate from the intrinsic change of cement and the physical/ chemical environmental conditions. Thus it is necessary to evaluate the durability of a concrete for the increase of reliability and safety of the whole system during the designed life time. Considering the dry storage system of spent fuel is the item which can create a lot of added value, the development of a dry storage cask is usually initiated by private enterprises among developed countries. The detail research results and specific design criteria for the safety assessment of a concrete cask have not been revealed to the public well. In this paper, the major expected degradation factors and related degradation models of concrete casks were investigated as part of the safety assessment by taking account of the site where Korea industrial nuclear power plants are located

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...

Surface storage of vitrified high-level radioactive waste in reinforced-concrete casks

International Nuclear Information System (INIS)

Beale, H.; George, M.W.; Robertson, T.J.M.

1982-06-01

The feasibility of storing canisters containing vitrified high level radioactive waste in reinforced concrete casks is examined. This preliminary study identifies the limitations and probable cost of such a store and leads to the conclusion that the concept is feasible. (author)

A Cask Processing Enclosure for the TRU Waste Processing Center - 13408

Energy Technology Data Exchange (ETDEWEB)

Newman, John T.; Mendez, Nicholas [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)

2013-07-01

This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

Design of double containment canister cask storage system

International Nuclear Information System (INIS)

Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

2004-01-01

Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

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

Science.gov (United States)

2013-12-06

... Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory Commission. ACTION... storage regulations by revising the Holtec International HI- STORM 100 Cask System listing within the...C) No. 1014. Amendment No. 9 broadens the subgrade requirements for the HI-STORM 100U part of the HI...

B cell remote-handled waste shipment cask alternatives study; TOPICAL

International Nuclear Information System (INIS)

RIDDELLE, J.G.

1999-01-01

The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

Cask system maintenance in the Federal Waste Management System

International Nuclear Information System (INIS)

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

1991-01-01

In early 1988, in support of the development of the transportation system for the Office of Civilian Radioactive Waste Management System (OCRWM), a feasibility study was undertaken to define a the concept for a stand-alone, ''green-field'' facility for maintaining the Federal Waste Management System (FWMS) casks. This study provided and initial layout facility design, an estimate of the construction costs, and an acquisition schedule for a Cask Maintenance Facility (CMF). It also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs derived from the study have been organized for use in the total transportation system decision-making process. Most importantly, they also provide a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone, ''green-field'' configuration. This design approach provides a comprehensive design evaluation, to guide the development of a cost estimate and to permit flexibility in locating the facility. The following sections provide background information on cask system maintenance, briefly summarizes some of the functional requirements that a CMF must satisfy, provides a physical description of the CMF, briefly discusses the cost and schedule estimates and then reviews the findings of the efforts undertaken since the feasibility study was completed. 15 refs., 3 figs

Final version dry cask storage study

International Nuclear Information System (INIS)

1989-02-01

This report was prepared in response to Section 5064 of the Nuclear Waste Policy Amendments Act of 1987 (the Amendments Act--Public Law 100-203), which directs the Secretary of Energy to conduct a study of the use of dry-cask-storage technology for storing spent fuel at the sites of civilian nuclear reactors until a geologic repository is available. In conducting this study, whose results are being reported to the Congress, the Secretary was to consider such factors as costs, effects on human health and the environment, and the extent to which the Nuclear Waste Fund can and should be used to provide funds for at-reactor storage. In addition, the Secretary was to consult with the Nuclear Regulatory Commission (NRC), include NRC comments in the report, and solicit the views of State and local governments and the public. The study performed in response to these requirements was based largely on data published by the DOE or the NRC or included in documents issued by the DOE. Among the DOE documents are the 1987 MRS proposal to the Congress and a subsequent report, prepared to supply the Congress with additional information on the MRS facility. Because in evaluating dry storage at reactor sites it is necessary to take into account other options for meeting storage needs, this study covered all forms of at-reactor storage. 107 refs., 15 figs., 10 tabs

Storage and transport casks combine to bring benefits

International Nuclear Information System (INIS)

Thorup, C.

1988-01-01

The Nuclear Assurance Corporation is currently preparing a safety report on its new spent fuel storage/transport casks. The report is due to be submitted to the NRC in 1989, together with an application for a licence. The aim of the combined casks is to simplify the process of dealing with spent fuel, whilst keeping costs down. The design of the casks is described, together with questions relating to the licensing of the casks. (author)

Safety analysis of spent fuel transport and storage casks under extreme impact conditions

International Nuclear Information System (INIS)

Wolff, D.; Wieser, G.; Ballheimer, V.; Voelzke, H.; Droste, B.

2005-01-01

Full text: Worldwide the security of transport and storage of spent fuel with respect to terrorism threats is a matter of concern. In Germany a spent nuclear fuel management program was developed by the government including a new concept of dry on-site interim storage instead of centralized interim storage. In order to minimize transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities, the operators of NPPs have to erect and to use interim storage facilities for spent nuclear fuel on the site or in the vicinity of nuclear power plants. Up to now, 11 on-site interim storage buildings, one storage tunnel and 4 on-site interim storage areas (preliminary cask storage till the on-site interim storage building is completed) have been licensed at 12 nuclear power plant sites. Inside the interim storage buildings the casks are kept in upright position, whereas at the preliminary interim storage areas horizontal storage of the casks on concrete slabs is used and each cask is covered by concrete elements. Storage buildings and concrete elements are designed only for gamma and neutron radiation shielding reasons and as weather protection. Therefore the security of spent fuel inside a dual purpose transport and storage cask depends on the inherent safety of the cask itself. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. Since the terror attacks of 11 September 2001 the determination of casks' inherent safety also under extreme impact conditions due to terrorist attacks has been of our increasing interest. With respect to spent fuel storage one of the most critical scenarios of a terrorist attack for a cask is the centric impact of a dynamic load onto the lid-seal-system caused e.g. by direct aircraft crash or its engine as well as by a

Performance of CASTORR HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

International Nuclear Information System (INIS)

Wilmsmeier, Marco; Vossnacke, Andre

2008-01-01

On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR R HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR R HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR R HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR R HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out as well. GNS long-term CASTOR R

Management plan for the procurement of shipping casks required to service proposed federal waste repositories

International Nuclear Information System (INIS)

Renken, J.H.; Dupree, S.A.; Allen, G.C. Jr.; Freedman, J.M.

1978-08-01

Development of transportation systems to move radioactive waste and unreprocessed spent fuel to proposed federal waste repositories is an integral part of the National Waste Terminal Storage Program. To meet this requirement, shipping casks must be designed, licensed, and fabricated. To assist the manager charged with this responsibility, a Cask Procurement Plan has been formulated. This plan is presented as a logic diagram that is suitable for computer analysis. In addition to the diagram, narrative material that describes various activities in the plan is also included. A preliminary computer analysis of the logic diagram indicates that, depending on the result of several decisions which must be made during the course of the work, the latest start dates which will allow prototype delivery of all types of casks by December 1985, range from November 1977 to March 1982

High capacity cask (TN28V) and International Transport System for the return shipment of vitrified high activity wastes

International Nuclear Information System (INIS)

Sert, G.; Savornin, B.; Rouquette, Y.

1989-01-01

The reprocessing of spent fuel generates different kinds of wastes. Among them fission products and non fissile actinides represent 98% of the radioactivity; these wastes are separated, concentrated, mixed with molten glass and poured into stainless steel containers. For political reasons, it is necessary to return these vitrified high activity wastes to the foreign countries which have decided to have their spent fuel reprocessed in France. So the transport of vitrified waste is vital for both the reprocessor and the utilities that have trusted the reprocessor and this operation has to be securely performed to give satisfaction to all concerned particles. For that reason Cogema will control the whole transport activity from La Hague plants to the receiving facilities of the customers. Therefore cogema will be responsible of the transport whatever the cask type (transport or storage) and will subcontract the transport operation to experienced companies such as Transnucleaire, PNTL or NTL, who will act on behalf of Cogema. Cogema will be the owner of the transport casks while the storage casks will normally be owned by the customers. Both cask types will of course have to comply with the requirements of La Hague, as published by Cogema

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.

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

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...

The dry spent RBMK fuel cask storage site at the Ignalina NPP in Lithuania

International Nuclear Information System (INIS)

Penkov, V.V.; Diersch, R.

1999-01-01

At present, there are about 15,000 spent RBMK fuel assemblies stored in the water pools near the reactors at the Ignalina Nuclear Power Plant (INPP). Part of them are cut in two bundles and stored in standardized baskets in the pools. Each basket is loaded with 102 bundles. For long-term interim storage of this fuel, it was decided to use dry storage in casks. For this reason, the total activity to be stored is split into individual units (casks). Each cask represents a closed and independent safety system, fulfilling all safety-relevant requirements for both normal operational and hypothetical accidental conditions. The main safety relevant features of the storage cask system are: (1) Inherent safety system; (2) Double barrier system; (3) Passive cooling by natural convection; (4) Safety against accidents. The cask dry storage system is a cost effective and multi-functional system for storage, transport after the operation time and final disposal under consideration of additional protective elements. From an economical point of view, cask storage has a number of advantages. Two cask types have been intended for the INPP storage site: (1) The CASTOR RBMK cask made of ductile cast iron; (2) The CONSTOR RBMK sandwich cask made of an inner and outer steel shell and reinforced heavy concrete. The CASTOR RBMK and the CONSTOR RBMK casks are designed to withstand severe storage site accidents and with help of impact limiters - to fulfil the IAEA test criteria for type B(U)F packages. The INPP spent RBMK fuel storage site is designed as an open air storage for an operational time of 50 years. The casks are arranged on the concrete storage pad. The site is equipped with a crane for cask handling and technological buildings and security systems. The safety analyses for fuel and cask handling and for cask handling and for cask technology at the site have been made and accepted by the Lithuanian Competent Authority. (author)

Fabrication and operational experience with the interim storage cask

International Nuclear Information System (INIS)

Scott, P.L.

1998-01-01

This paper discusses the fabrication and operational experience of the Interim Storage Cask (ISC). The ISC is a dry storage cask which is used to safely store a Core Component Container (CCC) containing up to seven Fast Flux Test Facility (FFTF) spent fuel assemblies at the US Department of Energy's Hanford Site. Under contract to B and W Hanford Company (BWHC), General Atomics (GA) designed and fabricated thirty ISC casks which BWHC is remotely loading at the FFTF facility. BWHC designed and fabricated the CCCS. As of December 1997, thirty ISCs have been fabricated, of which eighteen have been loaded and moved to a storage site adjacent to the FFTF facility. Fabrication consisted of three sets of casks. The first unit was completed and acceptance tested before any other units were fabricated. After the first unit passed all acceptance tests, nine more units were fabricated in the first production run. Before those nine units were completed, GA began a production run of twenty more units. The paper provides an overview of the cask design and discusses the problems encountered in fabrication, their resolution, and changes made in the fabrication processes to improve the quality of the casks. The paper also discusses the loading process and operational experiences with loading and handling of the casks. Information on loading times, worker dose exposure, and total dose for loading are presented

Shipping and storage cask data for spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Johnson, E.R.; Notz, K.J.

1988-11-01

This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask.

Shipping and storage cask data for spent nuclear fuel

International Nuclear Information System (INIS)

Johnson, E.R.; Notz, K.J.

1988-11-01

This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask

Multiple-Angle Muon Radiography of a Dry Storage Cask

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-23

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

Viability of Existing INL Facilities for Dry Storage Cask Handling

Energy Technology Data Exchange (ETDEWEB)

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

2013-04-30

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

Viability of Existing INL Facilities for Dry Storage Cask Handling

Energy Technology Data Exchange (ETDEWEB)

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

2013-04-01

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

Robotic radiation survey and analysis system for radiation waste casks

International Nuclear Information System (INIS)

Thunborg, S.

1987-01-01

Sandia National Laboratories (SNL) and the Hanford Engineering Development Laboratories have been involved in the development of remote systems technology concepts for handling defense high-level waste (DHLW) shipping casks at the waste repository. This effort was demonstrated the feasibility of using this technology for handling DHLW casks. These investigations have also shown that cask design can have a major effect on the feasibility of remote cask handling. Consequently, SNL has initiated a program to determine cask features necessary for robotic remote handling at the waste repository. The initial cask handling task selected for detailed investigation was the robotic radiation survey and analysis (RRSAS) task. In addition to determining the design features required for robotic cask handling, the RRSAS project contributes to the definition of techniques for random selection of swipe locations, the definition of robotic swipe parameters, force control techniques for robotic swipes, machine vision techniques for the location of objects in 3-D, repository robotic systems requirements, and repository data management system needs

Shipment and Storage Containers for Tritium Production Transportation Casks

International Nuclear Information System (INIS)

Massey, W.M.

1998-04-01

The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket

Moving the largest capacity PWR dual-purpose cask in the world from Goesgen NPP to the Zwilag interim storage site

International Nuclear Information System (INIS)

Delannay, M.; Dudragne, S.

2002-01-01

The Swiss Goesgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility of Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Goesgen NPP. Three transports of loaded TN24G casks between Goesgen and Zwilag were successfully performed at the beginning of 2002 with the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth transport of loaded TN24G was due to happen in October 2002. The TN24G cask, as part of the TN24 casks family, proved to be a very efficient solution for the KKG spent fuel management. (author)

A truck cask design for shipping defense high-level waste

International Nuclear Information System (INIS)

Madsen, M.M.; Zimmer, A.

1985-01-01

The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

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...

Development of concrete cask storage technology for spent nuclear fuel

International Nuclear Information System (INIS)

Saegusa, Toshiari; Shirai, Koji; Takeda, Hirofumi

2010-01-01

Need of spent fuel storage in Japan is estimated as 10,000 to 25,000 t by 2050 depending on reprocessing. Concrete cask storage is expected due to its economy and risk hedge for procurement. The CRIEPI executed verification tests using full-scale concrete casks. Heat removal performances in normal and accidental conditions were verified and analytical method for the normal condition was established. Shielding performance focus on radiation streaming through the air outlet was tested and confirmed to meet the design requirements. Structural integrity was verified in terms of fracture toughness of stainless steel canister for the cask of accidental drop tests. Cracking of cylindrical concrete container due to thermal stress was confirmed to maintain its integrity. Seismic tests of concrete cask without tie-down using scale and full-scale model casks were carried out to confirm that the casks do not tip-over and the spent fuel assembly keeps its integrity under severe earthquake conditions. Long-term integrity of concrete cask for 40 to 60 years is required. It was confirmed using a real concrete cask storing real spent fuel for 15 years. Stress corrosion cracking is serious issue for concrete cask storage in the salty air environment. The material factor was improved by using highly corrosion resistant stainless steel. The environmental factor was mitigated by the development of salt reduction technology. Estimate of surface salt concentration as a function of time became possible. Monitoring technology to detect accidental loss of containment of the canister by the stress corrosion cracking was developed. Spent fuel integrity during storage was evaluated in terms of hydrogen movement using spent fuel claddings stored for 20 years. The effect of hydrogen on the integrity of the cladding was found negligible. With these results, information necessary for real service of concrete cask was almost prepared. Remaining subject is to develop more economical and rational

Status of spent fuel shipping cask development

International Nuclear Information System (INIS)

Hall, I.K.; Hinschberger, S.T.

1989-01-01

This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

Burnup credit for storage and transportation casks

International Nuclear Information System (INIS)

Wells, A.H.

1988-01-01

The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety

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...

Design assessment for transport and storage casks

International Nuclear Information System (INIS)

Janberg, K.; Diersch, R.; Spilker, H.; Dreier, G.

1995-01-01

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress-strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved. (author)

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...

An economic evaluation of a storage system for casks with burnup credit

International Nuclear Information System (INIS)

Mimura, Masahiro; Tsuda, Kazuaki; Yamada, Nobuyuki; O-iwa, Akio.

1993-01-01

It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

CASTORR 1000/19: Development and Design of a New Transport and Storage Cask

International Nuclear Information System (INIS)

Funke, Th.; Henig, Ch.

2008-01-01

The design of the new transport and storage cask type CASTOR R 1000/19 is presented in this paper. This cask was developed for the dry interim storage of spent VVER1000 fuel assemblies concerning the requirements of the Temelin NPP, Czech Republic. While the cask body is based on well-known ductile cast iron cask types with in-wall moderator, the basket follows a new concept. The basket is able to carry 19 fuel assemblies with a total decay heat power up to approximately 17 kW. The cask fulfils all requirements for a type B(U)F package. The main nuclear, mechanical and thermal properties of the cask are illustrated for normal conditions and for hypothetical accident scenarios during transport and storage. The main steps of the handling procedure such as loading the cask, drying the cavity and mounting the double lid system for tightness during interim storage are shown in principle. For this handling, boundary conditions at the NPP site such as dimensions, weight and the loading machine interface are considered. (authors)

Drop test of reinforced concrete slab onto storage cask

International Nuclear Information System (INIS)

Kato, Y.; Hattori, S.; Ito, C.; Sirai, K.; Ozaki, S.; Kato, O.

1993-01-01

In this research, drop tests onto full-scale casks considering the specifications of a falling object (weight, construction, drop height, etc.) demonstrate and evaluate the integrity of casks in case a heavy object drops into the storage facilities. (J.P.N.)

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)

Signatures of Extended Storage of Used Nuclear Fuel in Casks

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-28

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

CASTOR THTR transport/storage casks

International Nuclear Information System (INIS)

Laug, R.W.; Spilker, H.; Sappok, M.

1998-01-01

For the management of spent fuel from nuclear power plants, two possibilities are available in Germany. One possibility is the reprocessing of the spent fuel and the realization of a so called closed nuclear fuel cycle, the other is the direct disposal after a period of interim storage, without reprocessing. For the German GCR plants ''THTR 300'' and ''AVR'', only the way of direct disposal is available to date for managing the spent fuel (pebble-bed fuel). For the period of interim storage, dry storage in casks was selected. (author)

Life cycle cost report of VHLW cask

International Nuclear Information System (INIS)

1995-06-01

This document, the Life Cycle Cost Report (LCCR) for the VHLW Cask, presents the life cycle costs for acquiring, using, and disposing of the VHLW casks. The VHLW cask consists of a ductile iron cask body, called the shielding insert, which is used for storage and transportation, and ultimately for disposal of Defense High Level Waste which has been vitrified and placed into VHLW canisters. Each ductile iron VHLW shielding insert holds one VHLW canister. For transportation, the shielding insert is placed into a containment overpack. The VHLW cask as configured for transportation is a legal weight truck cask which will be licensed by NRC. The purpose of this LCCR is to present the development of the life cycle costs for using the VHLW cask to transport VHLW canisters from the generating sites to a disposal site. Life cycle costs include the cost of acquiring, operating, maintaining, and ultimately dispositioning the VHLW cask and its associated hardware. This report summarizes costs associated with transportation of the VHLW casks. Costs are developed on the basis of expected usage, anticipated source and destination locations, and expected quantities of VHLW which must be transported. DOE overhead costs, such as the costs associated with source and destination facility handling of the VHLW, are not included. Also not included are costs exclusive to storage or disposal of the VHLW waste

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...

Study of shielding analysis methods for casks of spent fuel and radioactive waste

International Nuclear Information System (INIS)

Saito, Ai

2017-01-01

Casks are used for storage or transport spent fuels or radioactive waste. Because high shielding performances are required, it is very important to confirm the validity of shielding analysis methods in order to evaluate cask shielding abilities appropriately. For this purpose, following studies were carried out. 1) A series of parameter survey for several codes to evaluated the difference of the results. 2) Calculations using the MCNP code are effective and theoretically have better accuracy. However setting reasonable variance reduction parameters is indispensable. Therefore, effectiveness of the ADVANTG code which produces automatically reasonable variance reduction parameters is carried out by comparison with conventional method. As a result, the validity of shielding analysis methods for casks is confirmed. The results will be taken into consideration in our future shielding analysis. (author)

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

Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

International Nuclear Information System (INIS)

Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

2013-08-01

The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

Energy Technology Data Exchange (ETDEWEB)

Ibarra, Luis; Medina, Ricardo; Yang, Haori

2018-03-23

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

Performance of CASTOR{sup R} HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

Energy Technology Data Exchange (ETDEWEB)

Wilmsmeier, Marco; Vossnacke, Andre [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany)

2008-07-01

On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR{sup R} HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR{sup R} HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR{sup R} HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR{sup R} HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out

Thermal test and analysis of a spent fuel storage cask

International Nuclear Information System (INIS)

Yamakawa, H.; Gomi, Y.; Ozaki, S.; Kosaki, A.

1993-01-01

A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

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

Science.gov (United States)

Thiele, Holger; Börst, Frank-Michael

2017-09-01

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

SNF shipping cask shielding analysis

International Nuclear Information System (INIS)

Johnson, J.O.; Pace, J.V. III.

1996-01-01

The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

Combined Thermal Management and Power Generation Concept for the Spent Fuel Dry Storage Cask

International Nuclear Information System (INIS)

Kim, In Guk; Bang, In Cheol

2017-01-01

The management of the spent nuclear fuel generated by nuclear power plants is a major issue in Korea due to insufficient capacity of the wet storage pools. Therefore, it is considered that dry storage system is the one possible solution for storing spent fuel. A dual-purpose metal cask (transportation and storage) is currently developing in Korea. This cask has 21 of fuel assemblies and 16.8 kW of maximum decay heat. To evaluate the critical safety in normal/off normal and accident conditions, critical stabilities were conducted by using CSAS 6.0. The experimental investigation of heat removal of a concrete storage cask was also conducted under normal, off normal and accident conditions. The results of the evaluation showed a good safety of the dry storage cask. The results showed the enhanced thermal performance according to modification of flow rate. To verify combined thermal management and power generation concept, a new type of test facility for dry storage cask was designed in 1/10 scale of concrete dry storage cask. The experimental study involved the cooling methods that are an integrated system on the top of the dry cask and air flow path on the canister wall. The results showed the temperature distribution of the wall and inside of the dry cask at the normal condition. The influence of the change of the heat load and cooling system were investigated. The heat removal by the integrated system is approximately 20% of the total heat removal of the dry cask with reduced wall temperature. In these tests, economic analysis is conducted by applying the concept of the cost and efficiency. Under different decay power cases, the energy efficiency of the heat pipe and Stirling engine are determined and compared based on experimental results. The average efficiencies of the Stirling engine were the range of 2.375 to 3.247% under the power range of 35– 65W. These results showed that advanced dry storage concept had a better cooling performance in comparison with

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...

Thermal tests of a transport / Storage cask in buried conditions

International Nuclear Information System (INIS)

Yamakawa, H.; Gomi, Y.; Saegusa, T.; Ito, C.

1998-01-01

Thermal tests for a hypothetical accident which simulated accidents caused by building collapse in case of an earthquake were conducted using a full-scale dry type transport and storage cask (total heat load: 23 kW). The objectives of these tests were to clarify the heat transfer features of the buried cask under such accidents and the time limit for maintaining the thermal integrity of the cask. Moreover, thermal analyses of the test cask under the buried conditions were carried out on basis of experimental results to establish methodology for the thermal analysis. The characteristics of the test cask are described as well as the test method used. The heat transfer features of the buried cask under such accidents and a time for maintaining the thermal integrity of the cask have been obtained. (O.M.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

Noh, Jae Soo; Park, Younwon; Song, Sub Lee; Kim, Hyeun Min

2016-01-01

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

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...

Seismic Performance of Dry Casks Storage for Long- Term Exposure

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-30

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

Evaluation of canister weld flaw depth for concrete storage cask

Energy Technology Data Exchange (ETDEWEB)

Moon, Tae Chul; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Jung, Sung Hun; Lee, Young Oh; Jung, In Su [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of)

2017-03-15

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B and PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

Certifying the TN-BRP and TN-REG transportable storage demonstration casks

International Nuclear Information System (INIS)

Abbott, D.G.; Nolan, D.J.; Yoshimura, H.R.

1991-01-01

The US DOE has obtained US NRC certification to transport two transportable storage casks for a demonstration project. Because the casks had been built before the decision was made to obtain NRC certification, only limited modifications could be made to the casks. NRC's review resulted in several technical concerns that were subsequently resolved by design modifications, testing, and further analysis. Certification activities included qualifying the ferritic steel body material, modifying the borated stainless steel basket design, and extensive impact limiter testing. Recommendations for certifying future casks are presented based on experience with these casks

Next nuclear gamble: transportation and storage of nuclear waste

International Nuclear Information System (INIS)

Resnikoff, M.

1983-01-01

Accidents during transport of nuclear waste are more threatening - though less likely - than a reactor meltdown because transportation accidents could occur in the middle of a populous city, affecting more people and property than a plant accident, according to the Council on Economic Priorities, a non-profit public service research organization. Transportation, as presently practiced, is unsafe. Shipping containers, called casks, are poorly designed and constructed, CEP says. The problem needs attention because the number of casks filled with nuclear waste on the nation's highways could increase a hundred times during the next 15 years under the Nuclear Waste Policy Act of 1982, which calls for storage areas. Recommendations, both technical and regulatory, for reducing the risks are presented

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

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...

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Preliminary safety analysis of criticality for dual-purpose metal cask under dry storage conditions in South Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Taeman, E-mail: tmkim@korad.or.kr [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Dho, Hoseog; Baeg, Chang-Yeal [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Lee, Gang-uk [Korea Nuclear Engineering and Service Co. (KONES), Hyundai Plaza, 341-4 Jangdae-dong, Yuseong-gu, Daejeon (Korea, Republic of)

2014-10-15

Highlights: • DPC is under development led by Korea Radioactive Waste Agency in South Korea. • The results of criticality analysis with respect to design requirements. • The k{sub eff} under normal and off-normal conditions were 0.36 and 0.46, respectively. • In addition, the k{sub eff} under a postulated accident condition was evaluated to be 0.94. - Abstract: A dual-purpose metal cask is under development led by Korea Radioactive Waste Agency (KORAD) in Korea, for the dry interim storage and long-distance transportation. This cask comprises a main body made of carbon steel and a stainless steel Dry Shielded Canister (DSC), with stainless steel baskets inside to contain spent fuel assemblies. In this study, nuclear criticality safety analysis was conducted as a part of safety assessment of the metal cask. Analysis to show criticality safety in accordance with regulatory requirements of PWR spent fuel storage was carried out. 10CFR72.124 “Criteria for nuclear criticality safety” and the Regulatory Guide of the American Nuclear Society, ANSI/ANS-57.9 “Design Criteria for an Independent Spent Fuel” and US NRC's “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility” were employed as regulatory standard and criteria. This paper shows results of criticality analysis with respect to each designated criterion with modeling of a virtual nuclear fuel assembly and a cask body that induces the maximum reactivity among various design basis fuels of the metal cask. In addition, the sensitivity analysis of nuclear criticality taking into account the various modeling deviation such as manufacturing tolerance and modeling assumptions of conventional models was carried out to ensure the reliability of the analysis result. The criticality evaluation result of the metal cask and the maximum k{sub eff} under normal and off-normal conditions were 0.36884 and 0.46255, respectively. The maximum k{sub eff} under a postulated

Verification of heat removal capability of a concrete cask system for spent fuel storage

International Nuclear Information System (INIS)

Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatugu

2001-01-01

The reprocessing works comprising of a center of nuclear fuel cycle in Japan is now under construction at Rokkasho-mura in Aomori prefecture, which is to be operated in 2005. However, as reprocessing capacity of the works is under total forming amount of spent nuclear fuels, it has been essential to construct a new facility intermediately to store them at a period before reprocessing them because of prediction to reach limit of pool storage in nuclear power stations. There are some intermediate storage methods, which are water pool method for wet storage, and bolt method, metal cask method, silo method and concrete cask method for dry storage. Among many methods, the dry storage is focussed at a standpoint of its operability and economy, the concrete cask method which has a lot of using results in U.S.A. has been focussed as a method expectable in its cost reduction effect among it. The Ishikawajima-Harima Heavy Industries Co., Ltd. produced, in trial, a concrete cask with real size to confirm productivity when advancing design work on concrete cask. By using the trial product, a heat removal test mainly focussing temperature of concrete in the cask was carried out to confirm heat conductive performances of the cask. And, analysis of heat conductivity was also carried out to verify validity of its analysis model. (G.K.)

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...

CASTOR{sup R} 1000/19: Development and Design of a New Transport and Storage Cask

Energy Technology Data Exchange (ETDEWEB)

Funke, Th.; Henig, Ch. [GNS mbH, Hollestrasse 7A, 45127 Essen (Germany)

2008-07-01

The design of the new transport and storage cask type CASTOR{sup R} 1000/19 is presented in this paper. This cask was developed for the dry interim storage of spent VVER1000 fuel assemblies concerning the requirements of the Temelin NPP, Czech Republic. While the cask body is based on well-known ductile cast iron cask types with in-wall moderator, the basket follows a new concept. The basket is able to carry 19 fuel assemblies with a total decay heat power up to approximately 17 kW. The cask fulfils all requirements for a type B(U)F package. The main nuclear, mechanical and thermal properties of the cask are illustrated for normal conditions and for hypothetical accident scenarios during transport and storage. The main steps of the handling procedure such as loading the cask, drying the cavity and mounting the double lid system for tightness during interim storage are shown in principle. For this handling, boundary conditions at the NPP site such as dimensions, weight and the loading machine interface are considered. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

None

2014-02-27

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

Decontamination of transport casks and of spent fuel storage facilities

International Nuclear Information System (INIS)

1990-06-01

The present document provides an analysis of the technical papers presented at the meeting as well as a summary of the panel discussion. Conclusions and Recommendations: The meeting agreed that the primary source of contamination of transport casks is the production of radioactive isotopes in nuclear fuel and activation products of fuel components in nuclear reactors. The type, amount of mechanism for the release of these isotopes depend on the reactor type and fuel handling process. The widespread use of pools for the storage and handling of fuel provides an easy path for the transfer of contamination. Control of pool water conditions is essential for limiting the spread of contamination. For plants where casks are immersed in pools for loading, the immersion times should be minimised. Casks should be designed for ease of decontamination. The meeting discussed the use of stainless steel and suitable paints for coating casks. Designers should consider the appropriate coating for specific applications. The use of pressurized water for decontamination is recommended whenever possible. A number of commercially available reagents exist for decontaminating cask external surfaces. More work, however, is needed to cope with Pressurized Water Reactor crud within casks. Leaking fuel should be identified and isolated before storage in pools. Basic studies of the uptake and release of contamination from cask surfaces should be initiated. Standardization of methods of contamination measurement and instrumentation should be instituted. Refs, figs and tabs

Status of the Beneficial Uses Shipping System cask (BUSS)

International Nuclear Information System (INIS)

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

1994-01-01

The Beneficial Uses Shipping System cask is a Type B packaging developed by Sandia National Laboratories for the U.S. Department of Energy. The cask is designed to transport special form radioactive source capsules (cesium chloride and strontium fluoride) produced by the Department of Energy's Hanford Waste Encapsulation and Storage Facility. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations. The cask prototype has been fabricated and Certificates of Compliance have been obtained

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...

Time to rethink nuclear waste storage

International Nuclear Information System (INIS)

Flynn, J.; Kasperson, R.; Kunreuther, H.; Slovic, P.

1992-01-01

The authors feel that given the levels of public opposition and distrust, congress should scrap the current nuclear waste storage program and reconsider the options. They observe that no compelling reason currently exists for siting a permanent repository at an early date. Technology developed in the past decade, especially dry-cask storage, provides assurance that wastes from commercial reactors can be stored safely for a lengthy period at current sites. In the longer term, reprocessing may reduce the volume of high-level wastes; storage elsewhere than in a geological repository may prove attractive; and experimental techniques such as transmutation - aimed at radically reducing the amount of time that wastes remain highly radioactive - could help solve the problem. In the meantime, the authors suggest that the US must begin a long-term effort to engage the public in a process of active collaboration. In doing so, the US has much to learn from other countries, where innovative approaches and techniques have began to establish public confidence

Numerical Simulation of the Thermal Performance of a Dry Storage Cask for Spent Nuclear Fuel

Directory of Open Access Journals (Sweden)

Heui-Yung Chang

2018-01-01

Full Text Available In this study, the heat flow characteristics and thermal performance of a dry storage cask were investigated via thermal flow experiments and a computational fluid dynamics (CFD simulation. The results indicate that there are many inner circulations in the flow channel of the cask (the channel width is 10 cm. These circulations affect the channel airflow efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures at the top concrete lid and the upper locations of the concrete cask are higher than those permitted by the design specification. The installation of the salt particle collection device has a limited negative effect on the thermal dissipation performance of the dry storage cask.

Development of on-site accident criteria for waste transfer casks

International Nuclear Information System (INIS)

Uldrich, E.D.

1989-01-01

Removal of radioactive waste must withstand the scrutiny of the public and various regulatory offices. Currently, there is no standard accident criteria or methodology for intra-site shipments at the Idaho National Engineering Laboratory (INEL). Since the radioactive waste transfer casks only carry material within the INEL site boundaries and are not used for normal over-the-road transport, the requirements of 10 CFR 71 Packaging and Transportation of Radioactive Material, do not provide suitable requirements for cask design or safety analyses. The objective is to develop realistically conservative accident scenarios consistent with the limited uses at the INEL for which the cask is approved

A cask maintenance facility feasibility study

International Nuclear Information System (INIS)

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

1989-01-01

The Oak Ridge National Laboratory (ORNL) is developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, ''green field'' facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrievable Storage (MRS) facility. The data, design, and estimated costs resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. Fleet servicing facility studies, operational studies from current cask system operators, a definition of the CMF system requirements, and the experience of others in the radioactive waste transportation field were used as a basis for the feasibility study. In addition, several cask handling facilities were visited to observe and discuss cask operations to establish the functions and methods of cask maintenance expected to be used in the facility. Finally, a peer review meeting was held at Oak Ridge, Tennessee in August, 1988, in which the assumptions, design, layout, and functions of the CMF were significantly refined. Attendees included representatives from industry, the repository and transportation operations

Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

Directory of Open Access Journals (Sweden)

Dyah Sulistyani Rahayu

2018-03-01

Full Text Available DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF. At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i

COBRA-SFS thermal analysis of a sealed storage cask for the Monitored Retrievable Storage of spent fuel

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.

1986-01-01

The COBRA-SFS (Spent Fuel Storage) computer code was used to predict temperature distributions in a concrete Sealed Storage Cask (SSC). This cask was designed for the Department of Energy in the Monitored Retrievable Storage (MRS) program for storage of spent fuel from commercial power operations. Analytical results were obtained for nominal operation of the SSC with spent fuel from 36 PWR fuel assemblies consolidated in 12 cylindrical canisters. Each canister generates 1650 W of thermal power. A parametric study was performed to assess the effects on cask thermal performance of thermal conductivity of the concrete, the fin material, and the amount of radial reinforcing steel bars (rebar). Seven different cases were modeled. The results of the COBRA-SFS analysis of the current cask design predict that the peak fuel cladding temperature in the SSC will not exceed the 37 0 C design limit for the maximum spent fuel load of 19.8 kW and a maximum expected ambient temperature of 37.8 0 C (100 0 F). The results of the parametric analyses illustrate the importance of material selection and design optimization with regard to the SSC thermal performance

European experience in transport/storage cask for vitrified residues

International Nuclear Information System (INIS)

Otton, Camille; Sicard, Damien

2007-01-01

Available in abstract form only. Full text of publication follows: Because of the evolution of burnup of spent fuel to be reprocessed, the high activity vitrified residues would not be transported in the existing cask designs. Therefore, TN International has decided in the late nineties to develop a brand new design of casks with optimized capacity able to store and transport the most active and hottest canisters: the TN TM 81 casks currently in use in Switzerland and the TN TM 85 cask which shall permit in the near future in Germany the storage and the transport of the most active vitrified residues defining a thermal power of 56 kW (kilowatts). The challenges for the TN TM 81 and TN TM 85 cask designs were that the geometry entry data were very restrictive and were combined with a fairly wide range set by the AREVA NC Specification relative to vitrified residue canister. The TN TM 81 and the TN TM 85 casks have been designed to fully anticipate shipment constraints of the present vitrified residue production. It also used the feedback of current shipments and the operational constraints and experience of receiving and shipping facilities. The casks had to fit as much as possible in the existing procedures for the already existing flasks such as the TN TM 28 cask and TS 28 V cask, all along the logistics chain of loading, unloading, transport and maintenance. (authors)

Facility handling and operational considerations with dry storage casks

International Nuclear Information System (INIS)

Moegling, J.; McCreery, P.N.

1982-09-01

The Tennessee Valley Authority, in conjunction with US DOE and Pacific Northwest Laboratory, is conducting the first US commercial demonstration of spent fuel storage in casks. The two casks selected for this study are the Castor Ic, on loan from Gesellschaft fur Nuklear Service of Essen, West Germany and the DOE supplied REA 2023, manufactured by Ridihalgh, Eggers, and Associates, of Columbus, Ohio. Preparations began in the spring of 1982. The casks are expected to be loaded with fuel at Brown's Ferry Nuclear Station early in 1984, and the test completed about two years later. NRC is issuing a two-year license for this test under 10 CFR 72

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...

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...

A Multi-function Cask for At-Reactor Storage of Short-Cooled Spent Fuel, Transport, and Disposal

International Nuclear Information System (INIS)

Forsberg, C.W.

2004-01-01

The spent nuclear fuel (SNF) system in the United States was designed with the assumptions that SNF would be stored for several years in an at-reactor pool and then transported to reprocessing plants for recovery of fissile materials, that security would not be a major issue, and that the SNF burnups would be low. The system has evolved into a once-through fuel cycle with high-burnup SNF, long-term storage at the reactor sites, and major requirements for safeguards and security. An alternative system is proposed to better meet these current requirements. The SNF is placed in multi-function casks with the casks used for at-reactor storage, transport, and repository disposal. The cask is the handling package, provides radiation shielding, and protects the SNF against accidents and assault. SNF assemblies are handled only once to minimize accident risks, maximize security and safeguards by minimizing access to SNF, and reduce costs. To maximize physical protection, the cask body is constructed of a cermet (oxide particles embedded in steel, the same class of materials used in tank armor) and contains no cooling channels or other penetrations that allow access to the SNF. To minimize pool storage of SNF, the cask is designed to accept short-cooled SNF. To maximize the capability of the cask to reject decay heat and to limit SNF temperatures from short-cooled SNF, the cask uses (1) natural circulation of inert gas mixtures inside the cask to transfer heat from the SNF to the cask body and (2) an overpack with external natural-circulation, liquid-cooled fins to transfer heat from the cask body to the atmosphere. This approach utilizes the entire cask body area for heat transfer to maximize heat removal rates-without any penetrations through the cask body that would reduce the physical protection capabilities of the cask body. After the SNF has cooled, the cooling overpack is removed. At the repository, the cask is placed in a corrosion-resistant overpack before disposal

Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

International Nuclear Information System (INIS)

Luk, V.K.; Spencer, B.W.; Shaukat, S.K.; Lam, I.P.; Dameron, R.A.

2003-01-01

Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

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...

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

Testing of Metal Cask and Concrete Cask

International Nuclear Information System (INIS)

Shirai, K.; Wataru, M.; Takeda, H.; Tani, J.; Arai, T.; Saegusa, T.

2015-01-01

In Japan, the first interim spent fuel storage facility (ISF) outside of nuclear power plant site in use of dual-purpose metal cask is being planned to start its commercial operation in 2012 in Mutsu city, Aomori prefecture. The CRIEPI (Central Research Institute of Electric Power Industry) has executed several study programs on demonstrative testing for interim storage of spent fuel, mainly related to metal cask and concrete cask storage technology to reflect in Japanese safety requirements for dry casks issued by NISA/METI (Nuclear and Industrial Safety Agency, Ministry of Economy and Trade Industry). On top of that, the Japan Nuclear Energy Safety Organization (JNES) has executed study programs on spent fuel integrity, etc. This paper introduces the summary of these research programs. (author)

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

Fuel-assembly behavior under dynamic impact loads due to dry-storage cask mishandling

International Nuclear Information System (INIS)

1991-07-01

Continued operation of nuclear power plants is contingent on the ability to provide adequate storage of spent fuel. Until recently, utilities have been able to maintain interim in-pool spent fuel storage. However, many facilities have reached their capacity and are now faced with shipping their spent fuel in dry casks to alternate storage facilities. The objective of this report is to provide estimates of the structural integrity of irradiated LWR fuel rods subjected to impact loads resulting from postulated cask handling accidents. This is accomplished in five stages: (1) Material properties for irradiated fuel are compiled for use in the structural analyses. (2) Results from parametric analyses of representative assembly designs are used to determine the most limiting case for end and side drop postulated handling accidents. (3) Detailed structural analysis results are presented for these critical designs. The detailed analyses include the coupling of assembly interaction with the cask and cask internals. (4) Criteria for both ultimate stress and brittle fracture failure modes of fuel rod cladding are established. (5) Safe cask handling drop height limits are computed based on items 2 through 4 above. 44 figs., 18 tabs

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)

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...

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

Science.gov (United States)

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

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Burnup credit effect on proposed cask payloads

International Nuclear Information System (INIS)

Hall, I.K.

1989-01-01

The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted

Loads imposed on dual purpose casks in German on-site-storage facilities for long term intermediate storage of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Wetzel, N.; Rabe, O. [TUeV NORD EnSys Hannover GmbH und Co. KG, Hanover (Germany)

2004-07-01

In accordance with recent changes of the atomic energy act and in order to secure reliable removal of spent fuel from the nuclear power plants' fuel storage ponds the German utilities filed license applications for a total of 12 onsite- storage facilities for spent fuel assemblies. By the end of 2003 the last of these storage facilities were licensed and are currently under construction. The first on-site-storage facility of that line became operational in late 2002. There are several design lines of storage facilities with different handling procedures or possible accident conditions. Short term interim storage facilities for a few casks are characterized by individual concrete hoods shielding the casks in horizontal position whereas long term intermediate storage facilities currently erected for large numbers of casks typically feature a condensed pattern of casks stored in upright position and massive structures of reinforced concrete. TUeV Hannover/Sachsen-Anhalt e. V. (now TUeV NORD EnSys Hannover GmbH and Co. KG) has been contracted as a body of independent experts for the assessment of all related safety requirements on behalf of the national licensing authority, the federal office for radiation protection (BfS).

Loads imposed on dual purpose casks in German on-site-storage facilities for long term intermediate storage of spent nuclear fuel

International Nuclear Information System (INIS)

Wetzel, N.; Rabe, O.

2004-01-01

In accordance with recent changes of the atomic energy act and in order to secure reliable removal of spent fuel from the nuclear power plants' fuel storage ponds the German utilities filed license applications for a total of 12 onsite- storage facilities for spent fuel assemblies. By the end of 2003 the last of these storage facilities were licensed and are currently under construction. The first on-site-storage facility of that line became operational in late 2002. There are several design lines of storage facilities with different handling procedures or possible accident conditions. Short term interim storage facilities for a few casks are characterized by individual concrete hoods shielding the casks in horizontal position whereas long term intermediate storage facilities currently erected for large numbers of casks typically feature a condensed pattern of casks stored in upright position and massive structures of reinforced concrete. TUeV Hannover/Sachsen-Anhalt e. V. (now TUeV NORD EnSys Hannover GmbH and Co. KG) has been contracted as a body of independent experts for the assessment of all related safety requirements on behalf of the national licensing authority, the federal office for radiation protection (BfS)

Monitored Retrievable Storage conceptual system study: cask-in-trench

International Nuclear Information System (INIS)

1983-11-01

This report provides a description of the Cask-in-Trench Storage Concept which meets a specified set of requirements; an estimate of the costs of construction, operation and decommissioning of the concept; the costs required to expand the facility throughput and storage capability; and the life cycle costs of the facility. 22 figures, 34 tables

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...

Spent fuel storage and transport cask decontamination and modification. An overview of management requirements and applications based on practical experience

International Nuclear Information System (INIS)

1999-04-01

A large increase in the number of casks required for transport and/or storage of spent fuel is forecast into the next century. The principal requirement will be for increased number of storage and dual purpose (transport/storage) casks for interim storage of spent fuel prior to reprocessing or permanent disposal in both on-site and off-site storage facilities. Through contact with radioactive materials spent fuel casks will be contaminated on both internal and external surfaces. In broad terms, cask contamination management can be defined by three components: minimisation, prevention and decontamination. This publication is a compilation of international experience with cask contamination problems and decontamination practices. The objective is to present current knowledge and experience as well as developments, trends and potential for new applications in this field. Furthermore, the report may assist in new design or modification of existing casks, cask handling systems and decontamination equipment

FFTF disposable solid waste cask

Energy Technology Data Exchange (ETDEWEB)

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

1983-01-01

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

FFTF disposable solid waste cask

International Nuclear Information System (INIS)

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

1983-01-01

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

Permitting plan for the high-level waste interim storage

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2009-01-01

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

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)

Status of US storage efforts

International Nuclear Information System (INIS)

Leasburg, R.H.

1984-01-01

Tasks involved in the implementation of the Nuclear Waste Policy Act are discussed. The need for speedy action on applications to deal with spent fuel storage problems is stressed. The problems faced by the Virginia Electric and Power Company, where full core discharge capability at the 1600-megawatt Surry power station is expected to be reached in early 1986, are reviewed. It is pointed out that although the Nuclear Waste Policy Act does not apply in this case, the problems illustrate the situation that may be faced after the Act is implemented. Problems involved in intro-utility transhipments and dry cask storage of spent fuel from Surry, including transportation ordinances at state and local levels and approval for the use of dry casks for storage, are reported. The suggestion that dry casks be used for interim storage and eventual transport to monitored retrievable storage facilities or permanent storage sites is considered. It is pointed out that data from a proposed 3-utility demonstration program of dry cask storage of consolidated fuels and the storage of fuels in air should give information applicable to the timely implementation of the Nuclear Waste Policy Act

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

Science.gov (United States)

2012-02-17

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

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

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

Containment performance of transportable storage casks at 9m drop test

Energy Technology Data Exchange (ETDEWEB)

Tobita, H. [Hitachi Zosen Corp., Osaka (Japan); Araki, K. [Hitachi Zosen Diesel and Engineering Co., Ltd., Tokyo (Japan)

2004-07-01

Spent fuel transportable storage casks usually have a double lid closure system, which consists of primary and secondary lids, and gaskets, to keep the containment function during transportation and storage, and to monitor a leakage or containment function during storage. Metal gasket is planning to be used not only during storage but transportation of both before and after storage. As metal gasket will degrade its containment function by creep during storage period of 50 years, relative displacement such as opening and slide displacement between the flange of the containment vessel and the lid should be restricted to a small range. To maintain the containment performance, we provisionally adopted the maximum opening limit of 0.1mm and the maximum slide displacement limit of 3.0mm in the full-scale cask design based on the report of the fundamental experiment on the metal gasket which examines the relation between leakage rate and sealing gap. The purpose of this study is to analyse the behaviour of the sealed parts (lid and vessel body) under 9m-drop impact test conditions and to establish some analytical method to evaluate this behaviour. In this study, the drop test of 1/3scale model of Hitz-B69 cask with the double lids closure system was carried out, the behaviours of the seal part were measured by displacement sensors, and they were compared with the result of the numerical analysis carried out separateley.

Standard casks for the transport of LWR spent fuel. Storage/transport casks for long cooled spent fuel

International Nuclear Information System (INIS)

Blum, P.; Sert, G.; Gagnon, R.

1983-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks are presently used for European and intercontinental transports and manufactured under TRANSNUCLEAIRE supervision in different countries. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardization faciliting fabrication, operation and spare part supply. Recently, TRANSNUCLEAIRE also developed a new generation of casks for the dry storage and occasional transport of LWR spent fuel which has been cooled for 5 years or 7 years in case of consolidated fuel rods. These casks have an optimum payload which takes into account the shielding requirements and the weight limitations at most sites. This paper deals more particularly with the TN 24 model which exists in 4 versions among which one for 24 PWR 900 fuel assemblies and another one for the consolidated fuel rods from 48 of same fuel assemblies

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...

Comparison of predicted versus measured dose rates for low-level radioactive waste cask shipments

International Nuclear Information System (INIS)

Macher, Martin S.

1992-01-01

Shippers of low-level radioactive waste must select casks which will provide sufficient shielding to keep dose rates below the federal limit of 10 mr/hr at 2 meters from the vehicle. Chem-Nuclear Systems, Inc. uses a cask selection methodology which is based on shielding analysis code predictions with an additional factor of safety applied to compensate for inhomogeneities in the waste, uncertainties in waste characterization, and inaccuracy in the calculational methods. This proven cask selection methodology is explained and suggested factors of safety are presented based on comparisons of predicted and measured dose rates. A safety factor of 2 is shown to be generally appropriate for relatively homogeneous waste and a safety factor of between 3 and 4 is shown to be generally appropriate for relatively inhomogeneous wastes. (author)

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

International Nuclear Information System (INIS)

Almomani, Belal; Lee, Sanghoon; Kang, Hyun Gook

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Shipment and Storage Containers for Tritium Production Transportation Casks

International Nuclear Information System (INIS)

Massey, W.M.

1998-01-01

A shipping and storage container for the Tritium production transportation casks may be required but requirements for protection of the irradiated rods and radioactive contamination have not been finalized. This report documents the various possibilities for the container depending on the final requirements

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.

Possible use of dual purpose dry storage casks for transportation and future storage of spent nuclear fuel from IRT-Sofia

International Nuclear Information System (INIS)

Manev, L.; Baltiyski, M.

2003-01-01

Objectives: The main objective of the present paper is related to one of the priority goals stipulated in Bulgarian Governmental Decision No.332 from May 17, 1999 - removal of SNF from IRT-Sofia site and its exporting for reprocessing and/or for temporary storage at Kozloduy NPP site. The variant of using dual purpose dry storage casks for transportation and future temporary storage of SNF from IRT-Sofia aims to find out a reasonable alternative of the existing till now variant for temporary SNF storage under water in the existing Kozloduy NPP Spent Fuel Storage Facility until its export for reprocessing. Results: Based on the given data for the condition of 73 Spent Nuclear Fuel Assemblies (SNFA) stored in the storage pool and technical data as well as data for available equipment and IRT-Sofia layout the following framework are specified: draft technical features of dual purpose dry storage casks and their overall dimensions; the suitability of the available equipment for safety and reliable performance of transportation and handling operations of assemblies from storage pool to dual purpose dry storage casks; the necessity of new equipment for performance of the above mentioned operations; Assemblies' transportation and handling operations are described; requirements to and conditions for future safety and reliable storage of SNFA loaded casks are determined. When selecting the technical solutions for safety assurance during performance of site handling operations of IRT-Sofia and for description of the exemplary casks the Effective Bulgarian Regulations are considered. The experience of other countries in performance of transfer and transportation of SNFA from such types of research reactors is taken into account. Also, Kozloduy NPP experience in SNF handling operations is taken into account. Conclusions: The Decision of Council of Minister for refurbishment of research reactor into a low power one and its future utilization for experimental and training

Constor steel concrete sandwich cask concept for transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Diersch, R.; Dreier, G.; Gluschke, K.; Zubkov, A.; Danilin, B.; Fromzel, V.

1998-01-01

A spent nuclear fuel transport and storage sandwich cask concept has been developed together with the Russian company CKTI. Special consideration was given to an economical and effective way of manufacturing by using conventional mechanical engineering technologies and common materials. The main objective of this development was to fabricate these casks in countries not having highly specialized industries. Nevertheless, this sandwich cask concept fulfills both the internationally valid IAEA criteria for transportation and the German criteria for long-term intermediate storage. The basic cask concept has been designed for adaptation to different spent fuel specifications as well as handling conditions in the NPP. Recently, adaptations have been made for spent fuel from the RBMK and VVER reactors, and also for BWR spent fuel. The analyses of nuclear and thermal behaviour as well as of strength according to IAEA examination requirements (9-m-drop, 1-m-pin drop, 800 deg. C-fire test) and of the behaviour during accident scenarios at the storage site (drop, fire, gas cloud explosion, side impact) were carried out by means of recognized calculation methods and programmes. In a special experimental programme, the mechanical and thermodynamic properties of heavy concrete were examined and the reference values required for safety analyses were determined. The results of the safety analysis after drop tests according to IAEA-regulations as well as after 1 m-drops at the storage site were confirmed by means of a test programme using a scale model. The fabrication technology has been tested with help of a half scale cask model. The model has been prefabricated in Russia and completed in Germany. It has been shown that the CONSTOR cask can be fabricated in an effective and economic way. (authors)

Testing of the dual slab verification detector for attended measurements of the BN-350 dry storage casks

Energy Technology Data Exchange (ETDEWEB)

Santi, Peter A [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory

2009-01-01

The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of {sup 3}He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

Interim dry cask storage of irradiated Fast Flux Test Facility fuel

International Nuclear Information System (INIS)

Scott, P.L.

1994-09-01

The Fast Flux Test Facility (FFTF), located at the US Department of Energy's (DOE'S) Hanford Site, is the largest, most modern, liquid metal-cooled test reactor in the world. This paper will give an overview of the FFTF Spent Fuel Off load project. Major discussion areas will address the status of the fuel off load project, including an overview of the fuel off load system and detailed discussion on the individual components that make up the dry cask storage portion of this system. These components consist of the Interim Storage Cask (ISC) and Core Component Container (CCC). This paper will also discuss the challenges that have been addressed in the evolution of this project

Behaviour of a spent fuel transport-storage cask during an airplane crash

International Nuclear Information System (INIS)

Malesys, P.

1994-01-01

TRANSNUCLEAIRE has got an order for the design and manufacturing of dual purpose, transport and storage, casks for spent fuel.An original item of the qualification of the design of this cask, for the storage aspect, is the necessity to demonstrate the resistance to an air crash.The typical case taken into account for design is the crash of a military fighter (F16) with a total mass of 14600kg and an impact speed of 150ms -1 . The demonstration of the ability of the cask to withstand this test is provided by both calculation and test.Two cases were considered. For the first one, the projectile hits the cask at the centre of the anti-crash lid. For the second one, it hits the cask in the plane of the closure system.The first step of the qualification is based on calculations performed with a code designed to study the effects of crashes. The aim of the calculations is, mainly, to determine the missile which has to be shot, and to select the worst orientation for the impact.To provide a full justification of the acceptability of the impact as concerned leaktightness, a test has been performed on a one-third scale model. It has shown that it was not altered by the impact.The paper provides a full description of the method of analysis, results of the numerical analysis, conclusion of the test and how the combination of calculation and test demonstrates the ability of the cask to withstand an airplane crash. ((orig.))

Seismic Response Analysis and Test of 1/8 Scale Model for a Spent Fuel Storage Cask

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Park, C. G.; Koo, G. H.; Seo, G. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yeom, S. H. [Chungnam Univ., Daejeon (Korea, Republic of); Choi, B. I.; Cho, Y. D. [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2005-07-15

The seismic response tests of a spent fuel dry storage cask model of 1/8 scale are performed for an typical 1940 El-centro and Kobe earthquakes. This report firstly focuses on the data generation by seismic response tests of a free standing storage cask model to check the overturing possibility of a storage cask and the slipping displacement on concrete slab bed. The variations in seismic load magnitude and cask/bed interface friction are considered in tests. The test results show that the model gives an overturning response for an extreme condition only. A FEM model is built for the test model of 1/8 scale spent fuel dry storage cask using available 3D contact conditions in ABAQUS/Explicit. Input load for this analysis is El-centro earthquake, and the friction coefficients are obtained from the test result. Penalty and kinematic contact methods of ABAQUS are used for a mechanical contact formulation. The analysis methods was verified with the rocking angle obtained by seismic response tests. The kinematic contact method with an adequate normal contact stiffness showed a good agreement with tests. Based on the established analysis method for 1/8 scale model, the seismic response analyses of a full scale model are performed for design and beyond design seismic loads.

Demonstration of cask transportation and dry storage of spent nuclear fuel

International Nuclear Information System (INIS)

Teer, B.R.; Clark, J.

1984-01-01

Nuclear Fuel Services, Inc. and the Department of Energy's Idaho Operations Office have signed a cost sharing contract to demonstrate dual purpose shipping and storage casks for spent nuclear fuel. Transnuclear, Inc. has been selected by NFS to design and supply two forged steel casks - one for 40 PWR assemblies from the Ginna reactor, the other for 85 BWR assemblies from the Big Rock Point reactor. The casks will be delivered to West Valley in mid-1985, loaded with the fuel assemblies and shipped by rail to the Idaho National Engineering Laboratory. The shipments will be made under a DOE Certificate of Compliance which will be issued based on reviews by Oak Ridge National Laboratory of Transnuclear's designs

Considerations applicable to the transportability of a transportable storage cask at the end of the storage period

International Nuclear Information System (INIS)

Sanders, T.L.; Ottinger, C.A.; Brimhall, J.L.; Creer, J.M.; Gilbert, E.R.; Jones, R.H.; McConnell, P.E.

1991-11-01

Additional spent fuel storage capacity is needed at many nuclear power plant sites where spent fuel storage pools have either reached or are expected to reach maximum capacities before spent fuel can be removed. This analysis examines certain aspects of Transportable Storage Casks (TSC) to assist in the determination of their feasibility as an option for at-reactor dry storage. Factors that can affect in-transport reliability include: the quality of design, development, and fabrication activities; the possibilities of damage or error during loading and closure; in-storage deterioration or unanticipated storage conditions; and the potential for loss of storage period monitoring/measurement data necessary for verifying the TSC fitness-for-transport. The reported effort utilizes a relative reliability comparison of TSCs to Transport-Only Casks (TOC) to identify and prioritize those issues and activities that are unique to TSCs. TSC system recommendations combine certain design and operational features, such as in-service monitoring, pretransport assessments, and conservation design assumptions, which when implemented and verified, should sufficiently ensure that the system will perform as intended in a later transport environment

Safety aspects of long-term dry interim storage of Type B spent fuel and high-level transport casks

International Nuclear Information System (INIS)

Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

2004-01-01

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

EBRII cask characterization measurements

International Nuclear Information System (INIS)

Haggard, D.L.; Brackenbush, L.W.

1996-01-01

This report describes the measurements performed to provide the radionuclide content and verify the stated mass of special nuclear material (SNM) in Experimental Breeder Reactor EBRII casks stored in Trench 1, Burial Ground 4C, 218-WAC 200 West Area. this information is needed to characterize the curie content of each cask and the total curies in the storage area. Gamma assay techniques typically employed for nondestructive assay (NDA) were used to determine the gamma-emitting isotopes in each cask, which were fission and activation products from the spent fuel. Passive neutron counting was selected to verify the stated plutonium content because the fission and activation products masked any gamma emissions from plutonium. The fast neutrons emitted by plutonium are highly penetrating and easily detected through several inches of shielding. A slab neutron detector containing five 3 He proportional counters was used to determine the neutron emission rates and estimate the mass of plutonium present. The measurements followed the methods and procedures routinely used for nuclear waste assay and safeguard measurements. The measured neutron yields confirmed the declared plutonium content for the fuel elements, with the exception of several casks that contained recycled plutonium or americium target material. In these casks, the 244 Cm content masked the neutron emissions from any plutonium. For these casks, the plutonium content was estimated by correlation with the 244 Cm neutron emissions

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

International Nuclear Information System (INIS)

Qu, Jianmin; Bazant, Zdenek; Jacobs, Laurence; Guimaraes, Maria

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-30

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

Development of an evaluation method for long-term sealability of the spent fuel storage cask

International Nuclear Information System (INIS)

Kato, Osamu; Ito, Chihiro; Saegusa, Toshiari

1996-01-01

One of the characteristics of the cask storage method of spent fuel is that containment of radioactive materials is assured by the storage cask itself. Thus, the seal structure of the cask is designed to have a highly reliable multi-barrier system using metallic gaskets instead of the conventional rubber gaskets. Although, it has been reported that the containment feature of the metallic gaskets is influenced by the plastic deformation and stress relaxation of the gaskets for a short-term usage, no research report has been published on the containment feature of the metallic gaskets for a long-term usage. In this paper, the stress relaxation features of the metallic gaskets is investigated which will directly influence the long-term sealability of the storage cask, at first. Next, the relationship between the temperature/time dependence of the plastic deformation and the containment features of the metallic gaskets. Finally, an evaluation method of the long-term sealability from experimental data of a short-term behavior of the metallic gaskets is proposed. (author)

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

International Nuclear Information System (INIS)

Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

2004-01-01

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

Calculation of radiation exposure of the environment of interim storage facilities for the dry storage of spent fuel in dual-purpose casks

Energy Technology Data Exchange (ETDEWEB)

Wortmann, B.; Stratmann, W. [STEAG Encotec GmbH, Essen (Germany)

2004-07-01

Acceptance problems in the public concerning the transport of spent nuclear fuel elements and a new political objective of the Federal Government have forced the German utilities to embark on on-site interim storage projects for the temporary storage of spent nuclear fuel elements. STEAG encotec GmbH, Essen, Germany, was awarded contracts for the conceptual planning including necessary shielding calculations for the majority of the 13 nuclear sites which opted for the dry storage concept. The capacity of the storage facilities ranges from 80 to 100 casks, according to the storage needs of the plants. The average dose rate at the surface of each cask was limited to 0.5 mSv/h, independent of the type of radiation. These new buildings should not significantly increase the exposure of the public to radiation already originating from the existing nuclear power plant. The layout of the storage building therefore has to ensure that additional target values of 10-20 iSv/y are not exceeded. These very low target values as well as the requirement to avoid high mechanical impacts to the casks in case of external events led to a thickness of walls and ceilings of between 1.2 m and 1.3 m. To remove the decay heat from the casks by natural convection sufficient cross sections of the air inlet and outlet ducts are required.

Thermoelectric Powered Wireless Sensors for Dry-Cask Storage

Science.gov (United States)

Carstens, Thomas Alan

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

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...

Development of the GA-4 and GA-9 legal weight truck spent fuel casks

International Nuclear Information System (INIS)

Grenier, R.M.; Meyer, R.J.; Mings, W.J.

1993-01-01

General Atomics (GA) has designed two new truck casks under contract to the U.S. Department of Energy as part of the Office of Civilian Radioactive Waste Management (OCRWM) Cask System Development Program. The GA-4 and GA-9 Casks, when licensed by the U.S. Nuclear Regulatory Commission, will transport intact spent fuel assemblies from commercial nuclear reactor sites to a monitored retrievable storage facility or permanent repository. (J.P.N.)

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

International Nuclear Information System (INIS)

Almomania, Belal; Kang, Hyun Gook; Lee, Sanghoon

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Cask technology program activities

International Nuclear Information System (INIS)

Allen, G.C. Jr.

1986-01-01

The civilian waste cask technology program consists of five major activities: (1) technical issue resolution directed toward NRC and DOT concerns, (2) system concept evaluations to determine the benefits of proposals made to DOE for transportation improvements, (3) applied technology and technical data tasks that provide independent information and enhance technology transfer between cask contractors, (4) standards development and code benchmarking that provide a service to DOE and cask contractors, and (5) testing to ensure the adequacy of cask designs. The program addresses broad issues that affect several cask development contractors and areas where independent technical input could enhance the Office of Civilian Radioactive Waste Management goals

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

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

International Nuclear Information System (INIS)

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

1992-04-01

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

Shipping cask demand associated with United States Government storage of commercial spent fuel

International Nuclear Information System (INIS)

Daling, P.M.; Engel, R.L.

1983-05-01

There were primarily two objectives of this study. The first was to develop estimates of the shipping cask fleet size that will be needed in the United States in the near future. These estimates were compared with current US spent fuel cask fleet size to determine its adequacy to provide the transportation services. The second objective was to develop estimates of the transportation costs associated with future movements of spent fuel. The results of this study were based on assumptions that were made prior to passage of the Nuclear Waste Policy Act of 1982 which authorizes the Department of Energy (DOE) to provide Federal Interim Storage of spent fuel from commercial reactors. The Act requires that the DOE is responsible for transportation of the fuel, although private industry is to provide these services. This paper examined the impacts of various spent fuel management strategies on spent fuel transportation hardware requirements and transportation costs. Conclusions related to optimization of the spent fuel transportation system can be drawn from the results of this study. The conclusions can be affected by changing the given set of assumptions used in this analysis. 3 tables

Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

Energy Technology Data Exchange (ETDEWEB)

Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

2006-08-15

Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

BWR - Spent Fuel Transport and Storage with the TNTM9/4 and TNTM24BH Casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2006-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., the Muehleberg Nuclear Power Plant owner, is involved in this process and has elected to store its BWR spent fuel in a new high capacity dual-purpose cask, the TNeTeM24BH from the COGEMA Logistics/TRANSNUCLEAR TN TM 24 family. The Muehleberg BWR spent fuels are transported by road in a medium size shuttle transport cask and then transferred to a heavy transport/storage cask (dry transfer) in the hot cell of ZWILAG site. For that purpose, COGEMA Logistics designed and supplied: - Two shuttle casks, TN TM 9/4, mainly devoted to transport of spent fuel from Muehleberg NPP to ZWILAG. Licensed according to IAEA 1996, the TN TM 9/4 is a 40 ton transport cask, for 7 BWR high bum-up spent fuel assemblies. - A series of new high capacity dual-purpose casks, TN TM 24BH, holding 69 BWR spent fuels. Two transport campaigns took place in 2003 and 2004. For each campaign, ten TN TM 9/4 round trips are performed, and one TN TM 24BH is loaded. 5 additional TN TM 24BH are being manufactured for BKW, and the next transport campaigns are scheduled from 2006. The TN TM 24BH high capacity dual purpose cask and the TN TM 9/4 transport cask characteristics and capabilities will then be detailed. (authors)

Design of dry cask storage for Serpong multi purpose reactor spent nuclear fuel

International Nuclear Information System (INIS)

Dyah Sulistyani Rahayu; Yuli Purwanto; Zainus Salimin

2018-01-01

The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF, 198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decay at a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg 2 . The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method based on SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister. The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 °C and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I 0 from the rack no 3, 8 and 10 are 434.307; 446

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

IAEA'S International Working Group on Integrated Transport and Storage Safety case for Dual Purpose Casks

International Nuclear Information System (INIS)

Kumano, Yumiko; Varley, Kasturi; ); Droste, Bernhard; Wolff, Dietmar; Hirose, Makoto; Harvey, John; Reiche, Ingo; McConnell, Paul

2014-01-01

Spent nuclear fuel is generated from the operation of nuclear reactors and it is imperative that it is safely managed following its removal from reactor cores. Reactor pools are usually designed based on the assumption that the fuel will be removed after a short period of time either for reprocessing, disposal, or further storage. As a result of storing higher burn-up fuel, significantly increased time-frame till disposal solutions are prepared, and delays in decisions on strategies for spent fuel management, the volume of spent fuel discharged from reactors which needs to be managed and stored is on the increase. Consequently, additional storage capacity is needed following the initial storage in reactor pools. Options for additional storage include wet storage or dry storage in a dedicated facility or in storage casks. One of these options is the use of a Dual Purpose Cask (DPC), which is a specially designed cask for both storage and transport. The management of spent fuel using a DPC generally involves on-site and off-site transportation before and after storage. Most countries require package design approval for the DPC to be transported. In addition, it is required in many countries to have a licence for storage of the spent fuel in the DPC or a licence for a storage facility that contains DPCs. Therefore, demonstration of compliance of the DPC with national and international transport regulations as well as with the storage requirements is necessary. In order to address this increasing need among Member States, the IAEA established an international working group in 2010 to develop a guidance for integrating safety cases for both storage and transport in a holistic manner. The working group consists of experts from regulatory bodies, Technical Support Organizations, operators for both transportation and storage, and research institutes. This activity is planned to be completed by 2013. Currently, a technical report has been drafted and is expected to be

Ultimate storage of radioactive wastes annual report, 1973

International Nuclear Information System (INIS)

The present report is a cooperative effort by the Company for Radiation and Environmental Research, Munich, and the Company for Nuclear Research, Karlsruhe, and provides a survey of work carried out during 1973 in the area of ultimate storage of radioactive wastes. Mining and construction works which were carried out in the Asse Salt Mine near Remlingen both underground as well as above the ground and which were used for repair, maintenance and expansion of the operation consistent with its future tasks are reported. The storage of low-level wastes at the 750 m level and also the test-oriented storage of medium-level waste materials at the 490 m level were carried out within the reporting period. Shielded storage casks S7V developed by the GfK were used for the first time in September for transporting 200 l iron-hooped drums filled with medium-level radioactive wastes to Asse, each shipment always containing seven drums. With two round-trips a week taking place between the Nuclear Research Center, Karlsruhe and the Asse II shaft installation, 14 drums were brought each week so that, by the end of the year, the quantity in storage amounted to a total of 233 drums. Further information is provided concerning the present status of research work in the fields of oromechanics, geology and hydrology as well as other findings. Further, storage techniques are discussed which are presently in the planning stage

Experimental and analytical study for demonstration program on shielding of casks for high-level wastes

International Nuclear Information System (INIS)

Ueki, K.; Nakazawa, M.; Hattorl, S.; Ozaki, S.; Tamaki, H.; Kadotani, H.; Ishizuka, T.; Ishikawa, S.

1993-01-01

The following remarks were obtained from the experiment and the DOT 3.5 and the MCNP analyses on the gamma ray and the neutron dose equivalent rates in the cask of interest. 1. The cask has thinner neutron shielding parts around the trunnions. Significant neutrons streaming around the trunnion parts was observed which was also cleared by the MCNP analysis for the 252 Cf source experiment. Accordingly, detailed neutron streaming calculations are required to evaluate the dose levels around the trunnions when loading the vitrified high-level wastes. 2. The room-scattered obstructive neutrons, mainly originating from the neutrons penetrating around the trunnions, at the top and the bottom of the cask are reduced significantly by preparing the water tank at the top and the water layer at the bottom. Therefore, a more accurate experiment is to be carried out in the future shielding experiment especially for neutrons. However, because the water tank and the layer do not exist in the actual high-level wastes transport cask, the experiment without the water tank and layer are not dispensable to demonstrate the transport conditions of the actual cask, too. 3. The gamma-ray and the neutron dose equivalent rate distributions obtained from the DOT 3.5 and the MCNP calculations, respectively, agreed closely with the measured values in the cask areas of interest. Accordingly, the DOT 3.5 code and the MCNP code with the NESX estimator can be employed not only for the shielding analysis of the future experiments, but also for making a safety analysis report of high-level wastes transport casks. (J.P.N.)

Neutronic and thermal hydraulic of dry cask storage systems

International Nuclear Information System (INIS)

Yavuz, U.

2000-01-01

Interim spent fuel storage systems must provide for the safe receipt, handling, retrieval and storage of spent nuclear fuel before reprocessing or disposal. In the context of achieving these objectives, the following features of the design are to be taken into consideration: to maintain fuel subcritical, to remove spent fuel residualheat, and to provide for radiation protection. These features in the design of a dry cask storage system were analyzed for normal operating conditions by employing COBRA-SFS, SCALE4.4 (ORIGEN, XSDOSE, CSAS6) codes. For a metal-shielded type storage system, appropriate designs, in accordance with safety assurance limits of IAEA, were obtained for spent fuel burned to 33000, 45000 and 55000 MW d/t and cooled for 5 and 10 years

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

Energy Technology Data Exchange (ETDEWEB)

Yang, Haori; Hayward, Jason; Can, Liao; Liu, Zhengzhi

2018-03-31

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

Spent fuel transport and storage system for NOK: The TN52L, TN97L, TN24 BHL and TN24 GB casks

International Nuclear Information System (INIS)

Wattez, L.; Verdier, A.; Monsigny, P.-A.

2007-01-01

NOK nuclear power plants in Switzerland, LEIBSTADT (KKL) BWR nuclear power plant and BEZNAU (KKB) PWR nuclear power plant have opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN52L and TN97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TN24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN52L and TN97L casks by the KKL and ZWILAG operators. In 2004, NOK also ordered three TN24 GB transport and storage casks for PWR fuel types. These casks are presently being manufactured. (author)

Development of metal cask for nuclear spent fuel

International Nuclear Information System (INIS)

Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

2001-01-01

It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

Technical issues affecting the transport of dual purpose casks

International Nuclear Information System (INIS)

Sanders, T.L.; Ottinger, C.A.; Brimhall, J.L.; Gilbert, E.R.; Jones, R.H.

1989-01-01

Approximately 60,000 metric tons of uranium (MTU) spent fuel will be discharged by the projected 2003 startup date of a federal disposal system. Of this, approximately 15,000 MTU will require storage outside existing or projected pool storage capabilities (Orvis et al., 1984). At-reactor dry storage of spent fuel, including vault, caisson, and cask systems, is being considered as an alternative to accommodate this excess fuel. Two dry storage cask concepts are among those under consideration. One involves placing spent fuel in storage-only casks (SOC) until a monitored retrievable storage (MRS) facility or repository is open, when the spent fuel would be transferred to a transport-only cask (TOC) for shipment. The second option, the dual purpose or transportable storage cask (TSC), is a system that would serve for both storage and later transport. To carry out its purpose, a TSC must be shipped directly from a storage facility to a disposal facility without first being opened to evaluate the cask or the fuel. To assure that both the fuel and the cask are in a transportable condition after 20 to 40 years of storage requires: (1) a definition of expected storage conditions; (2) an assessment of the impact of expected storage conditions on the reliability of the components and functions of the TSC during transport; and (3) the development of an overall TSC system design and operational strategy which ensures that TSC transport reliability compares to that of a transport-only cask. The later requirement is related to defining what appropriate design features, pre-shipment inspection, and/or alternative fuel and cask monitoring requirements are necessary during long-term storage to ensure the cask will meet transport performance requirements during later transport. 8 refs., 1 fig., 1 tab

A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

Energy Technology Data Exchange (ETDEWEB)

Yoon, Hyoungju; Park, Kwangheon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

2015-05-15

The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years.

Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

Energy Technology Data Exchange (ETDEWEB)

Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

2011-07-01

The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Tyacke, M.

1993-08-01

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

Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

International Nuclear Information System (INIS)

Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

1991-01-01

This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

Selection of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive wastes

International Nuclear Information System (INIS)

1984-04-01

The monitored retrievable storage (MRS) concepts considered are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. These concepts were compared primarily upon the relative performance of the storage units on seven criteria which together encompass the key considerations for selecting an MRS concept, namely their ability to satisfy the MRS mission requirements. These criteria were: safety and licensing; environmental impacts; socioeconomic impacts; siting requirements; cost; concept maturity; and flexibility. Evaluations of the candidate concepts indicate that all of the concepts could satisfactorily serve in an MRS facility. However, using the above criteria, the two concepts selected for further design studies are the concrete cask (primary concept), better entitled the sealed storage cask, and the field drywell (alternate concept). It was recognized that the transportable metal storage cask may be used to supplement at-reactor storage until such time as the repository or MRS becomes available. Consequently, a hybrid storage facility may be required (e.g., one using concrete casks or field drywells, with the capability of receiving and storing the transportable cask). Both the concrete cask and the field drywell concepts can easily accommodate the transportable cask. Further design efforts will ensure the compatibility of the MRS designs with the transportable cask

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...

An analysis of contingencies for making casks available for use during the early years of federal waste management system operations

International Nuclear Information System (INIS)

Johnson, P.E.; Pope, R.B.; Wankerl, M.W.; Joy, D.S.; Shappert, L.B.; Danese, F.L.; Best, R.E.; Schmid, S.

1992-01-01

This paper reports on a study which has been performed to examine the contingencies that could be pursued by the Department of energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

An analysis of contingencies for making casks available for use during the early years of Federal Waste Management System operations

International Nuclear Information System (INIS)

Johnson, P.E.; Joy, D.S.; Pope, R.B.; Shappert, L.B.; Wankerl, M.W.; Best, R.E.; Schmid, S.; Danese, F.L.

1992-01-01

A study has been performed to examine the contingencies that could be pursued by the Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) for shipping spent fuel beginning in 1998. OCRWM's current plan is to initiate operations using early production units of Initiative I truck and rail/barge casks that are presently being designed. Contingencies to this plan were considered in case some unforeseen event occurs that precludes the Initiative I casks from entering into service early in 1998 in sufficient quantities (both numbers and types) to satisfy DOE's shipping needs. Specifically, the study addressed the potential availability of cask systems, selected several cask usage scenarios, determined the requirements for casks under these scenarios, generically assessed different strategies for acquiring casks or the use of casks, and generically assessed cask fabrication capabilities. Issues concerning both domestic and foreign resources were addressed with a focus on the first five years of Federal Waste Management System (FWMS) operation

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...

Technical issues affecting the transport of dual purpose casks

International Nuclear Information System (INIS)

Sanders, T.L.; Ottinger, C.A.; Brimhall, J.L.; Gilbert, E.R.; Jones, R.H.

1989-01-01

Spent fuel storage pools at many nuclear reactors in the US have already or will soon be filled to maximum capacity. Approximately 50,000 metric tons of uranium (MTU) spent fuel will be discharged by the projected 2003 start-up date of a federal disposal system. Of this, approximately 6,000 MTU will require storage outside existing or projected pool storage capabilities (DOE, 1988). At-reactor dry storage of spent fuel, including vault, caisson, and cask systems, is being considered as an alternative to accommodate this excess fuel. Two dry storage cask concepts are among those under consideration. One involves placing spent fuel in storage-only casks (SOC) until a monitored retrievable storage (MRS) facility or repository is open when the spent fuel would be transferred to a transport-only cask (TOC) for shipment. The second option, the dual purpose or transportable storage cask (TSC), is a system that would serve for both storage and later transport without requiring the spent fuel to be unloaded. To carry out its purpose, a TSC must be shipped directly from a storage facility to a disposal facility without first being opened to evaluate the cask or the fuel. To assure that both the fuel and the cask are in a transportable condition after 20 to 40 years of storage requires: (1) a definition of expected storage conditions; (2) an assessment of the impact of expected storage conditions on the reliability of the components and functions of the TSC during transport; and (3) the development of an overall TSC system design and operational strategy which ensures that TSC transport reliability meets or exceeds that of a transport-only cask. The later requirement is related to defining what appropriate design features, pre-shipment inspections, and/or alternative fuel and cask monitoring requirements are necessary during long-term storage to ensure the cask will meet transport requirements during later transport

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...

The development of a transportable storage cask

International Nuclear Information System (INIS)

Stuart, I.F.

1991-01-01

There are a number of different technologies for implementing interim storage of spent fuel at reactor sites. It is generally accepted that, if possible, expanding the capacity of existing fuel pools through the installation of compact racks and the use of fuel rod consolidation are the most economical first steps. Once these have been carried out, other alternatives must be employed if further capacity expansion is required. It is not the purpose of this paper to discuss the relative economics of these alternatives, since under specific constraints and conditions each one can be shown to have an economic benefit. However, it is the reduction in plant operations, the minimising of radiation exposure, the inherent flexibility and corresponding overall favourable economics that have led to the development of the dual purpose storage and transport cask in the past few years. (author)

Combined storage system for LWR spent fuel and high-level waste

International Nuclear Information System (INIS)

Baxter, B.J.; Ganley, J.T.; Washington, J.A.

1983-01-01

The MODREX storage system consists of four basic elements: (1) the storage canister, (2) the storage module, (3) the storage cask, and (4) the transport cask. The storage canister is the heart of the system and, when used in combination with the module or either of the casks, allows the MODREX system to respond quickly to varied storage system requirements. The MODREX system can be used to hold either spent fuel assemblies or canistered solidified HLW. The ability to combine a basic storage canister with either a concrete module or a metal cask provides flexibility to meet a wide range of storage requirements. The spent fuel is stored in a dry, inert atmosphere, which essentially eliminates corrosion or deterioration of the cladding during extended storage periods. The storage canister and concrete storage module provide additional barriers against radioactivity release, enhancing long-term safety. Heat dissipation is passive, eliminating the need for additional emergency cooling systems or special redundancy. Modular, expandable construction permits minimum initial investment and capital carrying charges; additional capacity is built and paid for only as it is needed, retaining flexibility. 6 references, 2 figures, 1 table

Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility

International Nuclear Information System (INIS)

Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki

2007-01-01

The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)

Nondestructive Evaluation of the VSC-17 Cask

International Nuclear Information System (INIS)

Sheryl Morton; Al Carlson; Cecilia Hoffman; James Rivera; Phil Winston; Koji Shirai; Shin Takahashi; Masaharo Tanaka

2006-01-01

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

Loading, transport and storage of casks of the type CASTOR registered HAW28M in the frame of vitrified high-level waste repatriation from France

International Nuclear Information System (INIS)

Horn, Thomas; Graf, Wilhelm; Gosch-Warning, Michaela

2011-01-01

Until 2005 the German nuclear power plant operators have contracts with AREVA NC (former COGEMA) and NDA (former BNFL) concerning the reprocessing of spent fuel elements. The reprocessed and vitrified radioactive waste has to be repatriated to Germany. Due to the reprocessing of spent fuel elements with increased burnup and the repatriation after shorter cooling time the total activity and the Cm-244 content of the high-level-waste coquilles have increased since 2008. Consequently the heat output has increased to 2 kW/coquille. Therefore the new transport cask type CASTOR registered HAW28M was developed. The authors describe the design of the casks, the licensing according to the German transport regulations, loading procedures, radiation measurements and shipment completion. In autumn 2011 the repatriation of vitrified high-level waste from France is supposed to be completed with the transport of eleven CASTOR registered HAW28M.

Criticality effects of longitudinal gaps in poison for storage/transport casks

International Nuclear Information System (INIS)

Wells, A.H.

1985-01-01

A series of criticality calculations was performed with the AMPX/KENO system to determine the sensitivity of the NAC S/T cask 31 assembly basket, which is optimized for a design-basis fuel enrichment of 3.7 wt % 235 U, to axial gaps in the boron neutron poison. The results of these calculations show that axial gaps in the boron cause no statistically detectable change in k/sub eff/ until a minimum gap size is reached. The minimum gap size to change k/sub eff/ is dependent on the basket segment length, because a longer segment length results in fewer gaps for a given active fuel length. Longer segment lengths are less sensitive to gaps in the neutron poison. A typical segment length of 12 to 18 in. is projected for a casting of aluminum/boron alloy, indicating that axial gaps in the neutron poison of 1 in. would be acceptable. This gap thickness is much greater than the intersegment gap produced by modern casting techniques. The investigation described here demonstrated that an axial gap in neutron poison is acceptable for basket castings of large storage/transport casks. A precedent for such gaps is the NLI-6502 cask, so a cask basket with intersegment gaps should be licensable

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-06

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

Radioactive waste on-site storage alternative

International Nuclear Information System (INIS)

Dufrane, K.H.

1983-01-01

The first, most frequently evaluated approach for the large producer is the construction of a relatively expensive storage building. However, with the likely possibility that at least one disposal site will remain available and the building never used, such expenditures are difficult to justify. A low cost, but effective alternative, is the use of ''On-Site Storage Containers'' (OSSC) when and if required. Radwaste is only stored in the OSSC if a disposal site is not available. A small number of OSSC's would be purchased initially just to assure immediate access to storage. Only in the unlikely event of total disposal sites closure would additional OSSC's have to be obtained and even this is cost effective. With two or three months of storage available on site, production lead time is sufficient for the delivery of additional units at a rate faster than the waste can be produced. The recommended OSSC design would be sized and shielding optimized to meet the needs of the waste generator. Normally, this would duplicate the shipping containers (casks or vans) currently in use. The reinforced concrete design presented is suitable for outside storage, contains a leakproof polyethylene liner and has remote sampling capability. Licensing would be under 10CFR50.59 for interim storage with long-term storage under 10CFR30 not an impossibility. Cost comparisons of this approach vs. building construction show that for a typical reactor plant installation, the OSSC offers direct savings even under the worst case assumption that no disposal sites are available and the time value of money is zero

Project B-589, 300 Area transuranic waste interim storage project engineering study

International Nuclear Information System (INIS)

Greenhalgh, W.O.

1985-08-01

The purpose of the study was to look at various alternatives of taking newly generated, remote-handled transuranic waste (caisson waste) in the 300 Area, performing necessary transloading operations and preparing the waste for storage. The prepared waste would then be retrieved when the Waste Isolation Pilot Plant becomes operational and transshipped to the repository in New Mexico with a minimum of inspection and packaging. The scope of this study consisted of evaluating options for the transloading of the TRU wastes for shipment to a 200 Area storage site. Preconceptual design information furnished as part of the engineering study is listed below: produce a design for a clean, sealed waste canister; hot cell loadout system for the waste; in-cell loading or handling equipment; determine transshipment cask options; determine assay system requirements (optional); design or specify transport equipment required; provide a SARP cost estimate; determine operator training requirements; determine waste compaction equipment needs if desirable; develop a cost estimate and approximate schedule for a workable system option; and update the results presented in WHC Document TC-2025

Materials issues in cask development

International Nuclear Information System (INIS)

Chapman, R.L.; Sorenson, K.B.

1987-01-01

The Department of Energy Office of Civilian Radioactive Waste Management (DOE-OCRWM) is chartered by Congress under the Nuclear Waste Policy Act (NWPA) to build a permanent repository for commercial spent nuclear fuel and to provide a supporting transportation system. The OCRWM-sponsored From-Reactor Cask Systems Acquisition Program is developing a family of casks suitable for transporting commercial spent fuel. Phase I of the program is in the process of procuring cask designs for further development and eventual licensing. New materials will probably be proposed for various components of the cask system. This paper identifies potential new materials as a function of their use in the cask (containment, shielding, etc). To the extent that the identified materials are new (not yet qualified for their intended application), this paper identifies probable technical issues and development efforts which may be required to qualify the materials for uses in transportation casks

Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

International Nuclear Information System (INIS)

McKinnon, M.A.; Creer, J.M.

1988-10-01

REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

Verification of maximum impact force for interim storage cask for the Fast Flux Testing Facility

International Nuclear Information System (INIS)

Chen, W.W.; Chang, S.J.

1996-01-01

The objective of this paper is to perform an impact analysis of the Interim Storage Cask (ISC) of the Fast Flux Test Facility (FFTF) for a 4-ft end drop. The ISC is a concrete cask used to store spent nuclear fuels. The analysis is to justify the impact force calculated by General Atomics (General Atomics, 1994) using the ILMOD computer code. ILMOD determines the maximum force developed by the concrete crushing which occurs when the drop energy has been absorbed. The maximum force, multiplied by the dynamic load factor (DLF), was used to determine the maximum g-level on the cask during a 4-ft end drop accident onto the heavily reinforced FFTF Reactor Service Building's concrete surface. For the analysis, this surface was assumed to be unyielding and the cask absorbed all the drop energy. This conservative assumption simplified the modeling used to qualify the cask's structural integrity for this accident condition

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-01

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

Human factors engineering applications to the cask design activities of the Civilian Radioactive Waste Management Program

International Nuclear Information System (INIS)

Lake, W.H.; Peck, M. III

1993-01-01

The use of human factors engineering (HFE) in the design and use of spent fuel casks being developed for the Department of Energy's Civilian Radioactive Waste Management Program is addressed. The safety functions of cask systems are presented as background for HFE considerations. Because spent fuel casks are passive safety devices they could be subject to latent system failures due to human error. It is concluded that HFE should focus on operations and verifications tests, but should begin, to the extent possible, at the beginning of cask design. Use of HFE during design could serve to eliminate or preclude opportunity for human error

Three-dimensional finite element impact analysis of a nuclear waste truck cask

International Nuclear Information System (INIS)

Miller, J.D.

1985-01-01

This paper presents a three-dimensional finite element impact analysis of a hypothetical accident event for the preliminary design of a shipping cask which is used to transport radioactive waste by standard tractor-semitrailer truck. The nonlinear dynamic structural analysis code DYNA3D run on Sandia's Cray-1 computer was used to calculate the effects of the cask's closure-end impacting a rigid frictionless surface on an edge of its external impact limiter after a 30-foot fall. The center of gravity of the cask (made of 304 stainless steel and depleted uranium) was assumed to be directly above the impact point. An elastic-plastic material constitutive model was used to calculate the nonlinear response of the cask components to the transient loading. Interactive color graphics (PATRAN and MOVIE BYU) were used throughout the analysis, proving to be extremely helpful for generation and verification of the geometry and boundary conditions of the finite element model and for interpretation of the analysis results. Results from the calculations show the cask sustained large localized deformations. However, these were almost entirely confined to the impact limiters built into the cask. The closure sections were determined to remain intact, and leakage would not be expected after the event. As an example of a large three-dimensional finite element dynamic impact calculation, this analysis can serve as an excellent benchmark for computer aided design procedures

Cask technology program activities

International Nuclear Information System (INIS)

Allen, G.C. Jr.

1986-01-01

The civilian waste cask technology program consists of five major activities: Technical issue resolution directed toward NRC and DOT concerns; system concept evaluations to determine the benefits of proposals made to DOE for transportation improvements; applied technology and technical data tasks that provide independent information and enhance technology transfer between cask contractors; standards development and code benchmarking that provide a service to DOE and cask contractors; and testing to ensure the adequacy of cask designs. This paper addresses broad issues that affect several cask development contractors and areas where independent technical input could enhance OCRWM goals

Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations

International Nuclear Information System (INIS)

Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

1991-12-01

This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 x 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990

Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

International Nuclear Information System (INIS)

Richard, R.F.

1995-01-01

It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

CSER 95-014: Criticality storage category for K Basin spent cartridge filters in ECOROK casks at central waste complex

International Nuclear Information System (INIS)

Miller, E.M.

1996-01-01

This CSER justifies storing K Basin spent cartridge filters in ECOROK 25-11 casks with liners under the limits in an existing criticality prevention specification, CPS-SW-149-00002, Rev./Mod. B-1, because the worst case fissionable material inventory is less then 1/2 the CPS limit and the cask is larger than allowed containers, the cask and liner have adequate iron content to meet the CPS requirements, and the cask concrete walls are thick enough to isolate the fissionable contents of each cask from neutron interaction with fissionable material external to each cask

Status update of the BWR cask simulator

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-01

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

Cask system design guidance for robotic handling

International Nuclear Information System (INIS)

Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

1990-10-01

Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs

Disposal/storage container development experience

International Nuclear Information System (INIS)

Morrow, R.W. Jr.; Van Hoesen, S.D.; Fowler, E.; Barreira, D.G.; Emmett, R.W.

1988-01-01

Developmental work is currently underway at the Oak Ridge National Laboratory to design and manufacture a radioactive waste container suitable for both storage and disposal of radioactive wastes. The container is designed to fulfill the Department of Energy and Nuclear Regulatory Commission requirements for on-site storage, as well as the Nuclear Regulatory Commission's requirements for high integrity containers. The project also involves meeting the strict design and manufacturing ANSI/ASME NQA-1 guidelines. Special provisions of the container include a double containment system, with the inner barrier being corrosion resistant, the capability to monitor the internal cavity of the container, and off-gas venting capability. Further, yet related developmental work includes evaluating the cask for other varied uses, such as a processing cask, an ALARA shield, and even the possibility of Department of Transportation approval for an over-the-road transport cask

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

Development of a dry transport and storage cask for spent LWR fuel assemblies in Spain

International Nuclear Information System (INIS)

Melches, C.; Uriarte, A.; Espallardo, J.A.

1982-01-01

One of the advantages of the cask storage concept is its flexibility which makes it specially attractive in the case of the Spanish circumstances. For these reasons the Empresa Nacional del Uranio, S.A. (ENUSA), Junta de Energia Nuclear (JEN) and Equipos Nucleares, S.A. (ENSA) initiated in 1981 a joint program for the development of a prototype cask for the dry transport and storage of spent fuel assemblies. This program includes as main steps the analysis of the conceptual design, the detailed design and experimental tests, the fabrication of a prototype and its licencing and safety testing. The mentioned program, which started in the early 1981, is scheduled to be completed at the end of 1984

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

International Nuclear Information System (INIS)

Lundeen, J.E.

1994-01-01

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

Logistics management for storing multiple cask plug and remote handling systems in ITER

International Nuclear Information System (INIS)

Ventura, Rodrigo; Ferreira, João; Filip, Iulian; Vale, Alberto

2013-01-01

Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario

Logistics management for storing multiple cask plug and remote handling systems in ITER

Energy Technology Data Exchange (ETDEWEB)

Ventura, Rodrigo, E-mail: rodrigo.ventura@isr.ist.utl.pt [Laboratório de Robótica e Sistemas em Engenharia e Ciência – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Filip, Iulian, E-mail: ifilip@gmail.com [Faculty of Mechanical Engineering – Technical University Gheorghe Asachi of Iasi, 61 Dimitrie Mangeron Bldv., Iasi 700050 (Romania); Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

2013-10-15

Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario.

BWR-spent fuel transport and storage with the TN trademark 9/4 and TN trademark 24BH casks

International Nuclear Information System (INIS)

Wattez, L.; Marguerat, Y.; Hoesli, C.

2004-01-01

The Swiss Nuclear Utilities have started in 2001 to store spent fuel in dry metallic dual-purpose casks in ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd., as Muehleberg Nuclear Power Plant owner, is involved in this process and has selected to store its spent fuel, a new high capacity dual-purpose cask, the TN trademark 24BH. For the transport in a medium size cask, COGEMA LOGISTICS has developed a new cask, the TN trademark 9/4, to replace the NTL9 cask, which performed numerous transports of BWR spent fuel in the past decades. Licensed IAEA 1996, the TN trademark 9/4 is a 40 ton transport cask, for 7 BWR high burn-up spent fuel assemblies. The spent fuel assemblies can be transferred in the ZWILAG hot cell in the TN trademark 24BH cask. The first use of these casks took place in 2003. Ten TN trademark 9/4 transports were performed, and one TN trademark 24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN trademark 24BH high capacity dual purpose cask, the TN trademark 9/4 transport cask and describe in detail their characteristics and possibilities

ROCKING. A computer program for seismic response analysis of radioactive materials transport AND/OR storage casks

International Nuclear Information System (INIS)

Ikushima, Takeshi

1995-11-01

The computer program ROCKING has been developed for seismic response analysis, which includes rocking and sliding behavior, of radioactive materials transport and/or storage casks. Main features of ROCKING are as follows; (1) Cask is treated as a rigid body. (2) Rocking and sliding behavior are considered. (3) Impact forces are represented by the spring dashpot model located at impact points. (4) Friction force is calculated at interface between a cask and a floor. (5) Forces of wire ropes against tip-over work only as tensile loads. In the paper, the calculation model, the calculation equations, validity calculations and user's manual are shown. (author)

Development of the nuclear ship MUTSU spent fuel shipping cask

International Nuclear Information System (INIS)

Ishizuka, M.; Umeda, M.; Nawata, Y.; Sato, H.; Honami, M.; Nomura, T.; Ohashi, M.; Higashino, A.

1989-01-01

After the planned trial voyage (4700 MWD/MTU) of the nuclear ship MUTSU in 1990, her spent fuel assemblies, initially made of two types of enriched UO 2 (3.2wt% and 4.4wt%), will be transferred to the reprocessing plant soon after cooling down in the ship reactor for more than one year. For transportation, the MUTSU spent fuel shipping casks will be used. Prior to transportation to the reprocessing plant, the cooled spent fuel assemblies will be removed from the reactor to the shipping casks and housed at the spent fuel storage facility on site. In designing the MUTSU spent fuel shipping cask, considerations were given to make the leak-tightness and integrity of the cask confirmable during storage. The development of the cask and the storage function demonstration test were performed by Japan Atomic Energy Research Institute (JAERI) and Mitsubishi Heavy Industries, Ltd. (MHI). One prototype cask for the storage demonstration test and licensed thirty-five casks were manufactured between 1987 and 1988

Ventilation and air conditioning system in waste treatment and storage facilities

International Nuclear Information System (INIS)

Kinoshita, Hirotsugu; Sugawara, Kazushige.

1987-01-01

So far, the measures concerning the facilities for treating and storing radioactive wastes in nuclear fuel cycle in Japan were in the state which cannot be said to be sufficient. In order to cope with this situation, electric power companies constructed and operated radioactive waste concentration and volume reduction facilities, solid waste storing facilities for drums, high level solid waste storing facilities, spent fuel cask preserving facilities and so on successively in the premises of nuclear power stations, and for the wastes expected in future, the research and the construction plan of the facilities for treating and storing low, medium and high level wastes have been advanced. The ventilation and air conditioning system for these facilities is the important auxiliary system which has the mission of maintaining safe and pleasant environment in the facilities and lowering as far as possible the release of radioactive substances to outside. The outline of waste treatment and storage facilities is explained. The design condition, ventilation and air conditioning method, the features of respective waste treatment and storage facilities, and the problems for the future are described. Hereafter, mechanical ventilation system continues to be the main system, and filters become waste, while the exchange of filters is accompanied by the radiation exposure of workers. (Kako, I.)

Transport of Spent Nuclear Fuels, High and Intermediate Level Wastes: A Continuous Challenge

International Nuclear Information System (INIS)

Otton, C.; Blachet, L.

2009-01-01

For more than 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the used nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. In this presentation we will focus on the casks for the spent fuel, high level waste and intermediate level waste transportation. Answering to the constant evolution of the nuclear industry transport needs is a challenge that TN International faces routinely. Concerning the spent nuclear fuel transportation, TN International has developed in the early 80's a fleet of TN12 type casks fitted with several types of baskets able to safely transport all the spent fuel from the nuclear power plant or the research laboratories to AREVA La Hague plant. The current challenge is the design of a new transport cask generation taking into account the needs of the industry for the next 30 years. The replacement of the TN12 cask generation is to be scheduled as the regulations have changed and the fuel characteristics have evolved. The new generation of casks will take into account all the technical evolutions made during the TN12 thirty years of use. MOX spent fuel has now its dedicated cask: the TN112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in 2008 in the EDF nuclear power plant of Saint-Laurent-des-Eaux. Concerning the high level waste such as the La Hague vitrified residues a whole fleet of casks has been developed such as the TN 28 VT dedicated to transport, the TN81 and TN85 dedicated to transport and storage. These casks have permitted the

Study on concrete cask for practical use. Heat removal test under normal condition

International Nuclear Information System (INIS)

Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

2005-01-01

In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

Structural evaluation of spent nuclear fuel storage facilities under aircraft crash impact. Numerical study on evaluation of sealing performance of metal cask subjected to impact force

International Nuclear Information System (INIS)

Namba, Kosuke; Shirai, Koji; Saegusa, Toshiari

2008-01-01

A lot of safety evaluations on the important nuclear facilities against the aircraft crash have been reported in other countries. But the condition and the evaluation method to define impact force of aircraft crash have not been described clearly in the reports. In Japan, public concern with the safety evaluation against aircraft crash is increasing. It is important to make clear the behavior of the storage facilities installing the metal casks on impact loading due to aircraft crash. In this study, concerning crash between commercial aircraft and storage facility, impact analysis using dynamic analysis code LS-DYNA has been executed. The results showed that the storage facility was not completely destroyed. But the rigid aircraft engine may penetrate into the storage facility with local failure. Thus, we assumed the engine hit a metal cask in the storage facility and evaluated sealing performance of the metal cask under the impact loading. If the engine with 90m/s crashed the storage facility having concrete wall of 85cm in thickness, the remaining velocity became 60m/s after penetration. We calculated impact force of the engine with 60m/s crashing into the metal cask. Concerning the metal cask loaded the impact force, impact analysis was executed. We assumed two directions of impact force. One is vertical load and another is horizontal load against the cask. The result showed that plastic strain was not generated on flanges of the 1st lid and the sealing performance of the cask was maintained in each impact case. (author)

Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

International Nuclear Information System (INIS)

Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

2006-01-01

With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge: protection of people and environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, COGEMA LOGISTICS has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear producers through their operational choices. We will focus on the COGEMA LOGISTICS technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfill both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (COGEMA LOGISTICS, TRANSNUCLEAR, MAINCO, and LEMARECHAL CELESTIN) in order to manage transportation of liquid and solid waste towards interim or final storage sites. The paper has the following contents: About radioactive waste; - Radioactive waste classification; - High level activity waste and long-lived intermediate level waste; - Long-lived low level waste; - Short-lived low- and intermediate level waste; - Very low level waste; - The radioactive waste in nuclear fuel cycle; - Packaging design and

ASME codification of ductile cast iron cask for transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Saegusa, Toshiari; Arai, Taku

2012-01-01

The CRIEPI has been executing research and development on ductile cast iron cask for transport and storage of spent nuclear fuel in order to diversify options of the casks. Based on the research results, the CRIEPI proposed materials standards (Section II) and structural design standards (Section III) for the ductile cast iron cask to the authoritative and international ASME (American Society of Mechanical Engineers) Codes. For the Section II, the CRIEPI proposed the JIS G 5504 material with additional requirement prohibiting repair of cast body by welding, etc. as well as the ASTM A874 material to the Part A. In addition, the CRIEPI proposed design stress allowables, physical properties (thermal conductivity, modulus of elasticity, etc.), and external pressure chart to the Part D. For the Section III, the CRIEPI proposed a fracture toughness requirement of the ductile cast iron cask at -40degC to WB and WC of Division 3. Additionally, the CRIEPI proposed a design fatigue curve of the ductile cast iron cask to Appendix of Division 1. This report describes the outline of the proposed standards, their bases, and the deliberation process in order to promote proper usage of the code, future improvement, etc. (author)

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

Application of advanced handling techniques to transportation cask design

International Nuclear Information System (INIS)

Bennett, P.C.

1992-01-01

Sandia National Laboratories supports the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) applying technology to the safe transport of nuclear waste. Part of that development effort includes investigation of advanced handling technologies for automation of cask operations at nuclear waste receiving facilities. Although low radiation levels are expected near transport cask surfaces, cumulative occupational exposure at a receiving facility can be significant. Remote automated cask handling has the potential to reduce both the occupational exposure and the time necessary to process a cask. Thus, automated handling is consistent with DOE efforts to reduce the lifecycle costs of the waste disposal system and to maintain public and occupational radiological risks as low as reasonably achievable. This paper describes the development of advanced handling laboratory mock-ups and demonstrations for spent fuel casks. Utilizing the control enhancements described below, demonstrations have been carried out including cask location and identification, contact and non-contact surveys, impact limiter removal, tiedown release, uprighting, swing-free movement, gas sampling, and lid removal operations. Manually controlled movement around a cask under off-normal conditions has also been demonstrated

Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

International Nuclear Information System (INIS)

1988-01-01

The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study

International Nuclear Information System (INIS)

Tyacke, M.

1993-08-01

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

Impact analysis of spent fuel dry casks under accidental drop scenarios

International Nuclear Information System (INIS)

Braverman, J.I.; Morante, R.J.; Xu, J.; Hofmayer, C.H.; Shaukat, S.K.

2003-01-01

A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister (MPC) that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel. (author)

Test program of the drop tests with full scale and 1/2.5 scale models of spent nuclear fuel transport and storage cask

International Nuclear Information System (INIS)

Kuri, S.; Matsuoka, T.; Kishimoto, J.; Ishiko, D.; Saito, Y.; Kimura, T.

2004-01-01

MHI have been developing 5 types of spent nuclear fuel transport and storage cask (MSF cask fleet) as a cask line-up. In order to demonstrate their safety, a representative cask model for the cask fleet have been designed for drop test regulated in IAEA TS-R-1. The drop test with a full and a 1/2.5 scale models are to be performed. It describes the test program of the drop test and manufacturing process of the scale models used for the tests

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

International Nuclear Information System (INIS)

Saueressig, P.T.

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Saueressig, P.T.

1994-11-08

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

Improved bolt models for use in global analyses of storage and transportation casks subject to extra-regulatory loading

International Nuclear Information System (INIS)

Kalan, R.J.; Ammerman, D.J.; Gwinn, K.W.

2004-01-01

Transportation and storage casks subjected to extra-regulatory loadings may experience large stresses and strains in key structural components. One of the areas susceptible to these large stresses and strains is the bolted joint retaining any closure lid on an overpack or a canister. Modeling this joint accurately is necessary in evaluating the performance of the cask under extreme loading conditions. However, developing detailed models of a bolt in a large cask finite element model can dramatically increase the computational time, making the analysis prohibitive. Sandia National Laboratories used a series of calibrated, detailed, bolt finite element sub-models to develop a modified-beam bolt-model in order to examine the response of a storage cask and closure to severe accident loadings. The initial sub-models were calibrated for tension and shear loading using test data for large diameter bolts. Next, using the calibrated test model, sub-models of the actual joints were developed to obtain force-displacement curves and failure points for the bolted joint. These functions were used to develop a modified beam element representation of the bolted joint, which could be incorporated into the larger cask finite element model. This paper will address the modeling and assumptions used for the development of the initial calibration models, the joint sub-models and the modified beam model

Alternative cask maintenance facility concepts

International Nuclear Information System (INIS)

Attaway, C.R.; Pope, R.B.; Wiliamson, A.C.; Medley, L.G.; Shappert, L.B.

1992-01-01

In this paper, the results of three trade-off studies of alternative concepts for performing cask maintenance for Civilian Radioactive Waste Management System casks are presented. An earlier study resulted in a recommendation that a submerged pool concept for cask internal component removal be used in the design of a Cask Maintenance Facility. The first trade-off study resulted in confirming the previous recommendation that a submerged pool concept be used rather than an isolation cell; the basis for this continued recommendation is discussed. The second study provides an evaluation of the previously proposed facility for the capability of handling an increased quantity of OCRWM casks. The third study provides a preliminary concept for adding the capability to repaint the exterior cylindrical portions of casks

Evaluation of the impact behavior of the contents of reprocessing radioactive waste shipping cask subjected to drop impact

International Nuclear Information System (INIS)

Shirai, K.; Ito, C.; Funahashi, M.

1993-01-01

In this study, to investigate the impact response characteristics of the contents in the cask precisely, we performed the laboratory-scale drop tests, and on the basis of the test results, we proposed the construction of the spring-mass model and confirmed the accuracy of the proposed drop analysis method by comparison with drop test using a full-scale cask for high level wastes. Following the results of the drop tests and analysis, the outline of the contents and results is summarized below. 1) The drop tests onto the unyielding surface using a scale model containing several contents were performed and the effect of the interaction between the contents and the cask body on the impact response experimentally. 2) The above interaction can be characterized by the gap between the contents and the cask body caused by the release of the gravitational force at the moment the drop started. So, we proposed the analysis method for considering gap using spring-mass model by comparing the laboratory-scale drop test results. 3) We applied the proposed analysis method to a drop test using a full-scale cask for high level wastes, and it was found that this method seems to be good and convenient enough to evaluate the impact behavior of the contents in a transport cask subjected to drop impact. (J.P.N.)

Estimation of the heat generation in vitrified waste product and shield thickness of the cask for the transportation of vitrified waste product using Monte Carlo technique

International Nuclear Information System (INIS)

Deepa, A.K.; Jakhete, A.P.; Mehta, D.; Kaushik, C.P.

2011-01-01

High Level Liquid waste (HLW) generated during reprocessing of spent fuel contains most of the radioactivity present in the spent fuel resulting in the need for isolation and surveillance for extended period of time. Major components in HLW are the corrosion products, fission products such as 137 Cs, 90 Sr, 106 Ru, 144 Ce, 125 Sb etc, actinides and various chemicals used during reprocessing of spent fuel. Fresh HLW having an activity concentration of around 100Ci/l is to be vitrified into borosilicate glass and packed in canisters which are placed in S.S overpacks for better confinement. These overpacks contain around 0.7 Million Curies of activity. Characterisation of activity in HLW and activity profile of radionuclides for various cooling periods sets the base for the study. For transporting the vitrified waste product (VWP), two most important parameters is the shield thickness of the transportation cask and the heat generation in the waste product. This paper describes the methodology used in the estimation of lead thickness for the transportation cask using the Monte Carlo Technique. Heat generation due to decay of fission products results in the increase in temperature of the vitrified waste product during interim storage and disposal. Glass being the material, not having very high thermal conductivity, temperature difference between the canister and surrounding bears significance in view of the possibility of temperature based devitrification of VWP. The heat generation in the canister and the overpack containing vitrified glass is also estimated using MCNP. (author)

Modification of SKYSHINE-III to include cask array shadowing

Energy Technology Data Exchange (ETDEWEB)

Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States); Pfeifer, H.J. [NAC International, Norcross, GA (United States); Napolitano, D.G. [NISYS Corporation, Duluth, GA (United States)

2000-03-01

The NAC International version of SKYSHINE-III has been expanded to represent the radiation emissions from ISFSI (Interim Spent Fuel Storage Installations) dry storage casks using surface source descriptions. In addition, this modification includes a shadow shielding algorithm of the casks in the array. The resultant code is a flexible design tool which can be used to rapidly assess the impact of various cask loadings and arrangements. An example of its use in calculating dose rates for a 10x8 cask array is presented. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

2012-07-01

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National

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

Science.gov (United States)

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

2018-01-01

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

A cask maintenance facility feasibility study

International Nuclear Information System (INIS)

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

1989-01-01

The Oak Ridge National Laboratory (ORNL) is supporting the USDOE Office of Civilian Radioactive Waste Management (OCRWM) in developing a transportation system for spent nuclear fuel (SNF) and defense high level waste (HLW) as a part of the Federal Waste Management System (FWMS). In early 1988, a feasibility study was undertaken to design a stand-alone, green field facility for maintaining the FWMS casks. The feasibility study provided an initial layout facility design, an estimate of the construction cost, and an acquisition schedule for a Cask Maintenance Facility (CMF). The study also helped to define the interfaces between the transportation system and the waste generators, the repository, and a Monitored Retrieveable Storage (MRS) facility. The data, design, and estimated cost resulting from the study have been organized for use in the total transportation system decision-making process. Most importantly, the feasibility study also provides a foundation for continuing design and planning efforts. The feasibility study was based on an assumed stand-alone green field configuration because of the flexibility this design approach provides. A stand-alone facility requires the inclusion with support functions as well as the primary process facilities thus yielding a comprehensive design evaluation and cost estimate. For example, items such as roads, security and waste processing which might be shared with an integrated or collocated facility have been fully costed in the feasibility study. Thus, while the details of the facility design might change, the overall concept used in the study can be applied to other facility configurations as planning for the total FWMS develops

Swedish waste management

International Nuclear Information System (INIS)

Sandwall, L.

2004-01-01

Sweden has a well-functioning organization for managing various types of radioactive waste. There is an interim storage facility for spent nuclear fuel, a final repository for low and intermediate level waste, and a specially-built vessel with transport casks and containers for shipping the radioactive waste between the nuclear installations. (author)

A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

International Nuclear Information System (INIS)

Kim, Chang Hyun

1997-02-01

A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

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...

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...

Experience Of Using Metal-and-Concrete Cask TUK-108/1 For Storage And Transportation Of Spent Nuclear Fuel Of Decommissioned NPS

Energy Technology Data Exchange (ETDEWEB)

Barnes, E.; Dyer, R. [Environmental Protection Agency, Ronald Reagan Bldg. 3rd Floor 1200 Pennsylvania Av., NW Washington, D.C. 20024 (United States); Snipes, R. [Oak Ridge National Laboratories, VA (United States); Dolbenkov, V.G.; Guskov, V.D.; Korotkov, G.V. [Joint Stock Company ' KBSM' , 64 Lesnoy Av., St.Petersburg 194100 (Russian Federation); Makarchuk, T.F. [Joint Stock Company ' Atomstroyexport' , Potapovskiy str. 5, bld. 4, Moscow, 101990 (Russian Federation); Zakharchev, A.A. [State Corporation ' Rosatom' , 24-26 Ordinka St., Moscow, 100000 (Russian Federation)

2009-06-15

In past 10 years in Russia an intensive development of a new technology of management of spent nuclear fuel (SNF) has taken place. This technology is based on the concept of using a shielded cask which provides safety of its content (SNF) and meeting all other safety requirements to storage and transportation of SNF. Radiation protection against emission and non-propagation of activity outside the cask is ensured by the physical barriers such as all-metal or composite body, face work, inner structures to accommodate spent fuel assemblies (SFA), lids with sealing systems. Residual heat buildup is off-taken to the environment by natural way: emission and convection of surrounding air. The necessity in development of the cask technology of SNF management was conditioned by the situation at hand with defueling of Russian decommissioned nuclear-powered submarines (NPS) as the existed transport infrastructure and enterprises involved in fuel processing could not meet the demand for transportation and processing of SNF neither from reactors of all dismantled NPS, nor from reactors of NPS waiting for decommissioning. The US and Norway actively participated in the trilateral joint project with the Russian Federation aimed at creation of a cask prototype for interim storage and transportation of SNF of dismantled NPS. The 1.1 Project is a part of the Arctic Military Environmental Cooperation (AMEC) Program. In December 2000 the project was successfully completed by issuance of the certificate-permit for design and transportation of NP Submarine SNF. It was a first certified dual-purpose TUK from the MMC family. In these years 106 TUK-108/1 casks have been manufactured and supplied to PO Mayak, JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. The storage pads for interim storage of TUK-108/1 have been built and currently are in operation on sites of SNF unloading from submarine reactors and SNF cask-loading such as JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. In

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

A utility's perspective on the Nuclear Waste Policy Act

International Nuclear Information System (INIS)

Berry, W.W.

1985-01-01

The Nuclear Waste Policy Act is especially important to utilities because their customers pay for the disposal program, and the program is vital to nuclear operations and reconsideration of the nuclear option. DOE's accomplishments in implementing the Act are noteworthy, but we are concerned that some of them have been achieved later than specified by the schedule in the Act. We make recommendations regarding disposal fees, defense wastes, and shipping casks. Virginia Power has adopted a three-part strategy relying mainly on developing dry cask storage to solve the company's interim storage problems

Evaluation of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Triplett, M.B.; Smith, R.I.

1984-04-01

The primary mission selected by DOE for the monitored retrieval storage (MRS) system is to provide an alternative means of storage in the event that the repository program is delayed. The MRS concepts considered were the eight concepts included in the MRS Research and Development Report to Congress (DOE 1983). These concepts are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. Conceptual design analyses were performed for the candidate concepts using a common set of design requirements specified in consideration of the MRS mission

Report on the joint USA-Germany drop test program for a vitrified high level waste cask

International Nuclear Information System (INIS)

Golliher, K.G.; Witt, C.R.; Wieser, K.E.

1993-01-01

A series of full-scale drop tests was performed on a ductile iron transport cask in a cooperative program between the US Department of Energy (DOE) and Bundesantalt fuer Materialpruefung (BAM) in Germany. The tests, which were performed at BAM's test facility located near Lehre, Germany, were performed on a prototype cask designed for transport of Vitrified High Level Waste (VHLW) canisters. The VHLW cask is a right circular cylinder with a diameter of 1156 mm and a height of 3454 cm, and weighs approximately 24.6 kg including its payload of a single VHLW canister. The drop tests were performed with a non-radioactive, prototype VHLW canister in the cavity. (J.P.N.)

Techno-economical Analysis of High Level Waste Storage and Disposal Options

International Nuclear Information System (INIS)

Bace, M.; Trontl, K.; Vrankic, K.

2002-01-01

Global warming and instability of gas and oil prices are redefining the role of nuclear energy in electrical energy production. A production of high-level radioactive waste (HLW), during the nuclear power plant operation and a danger of high level waste mitigation to the environment are considered by the public as a main obstacle of accepting the nuclear option. As economical and technical aspects of the back end of fuel cycle will affect the nuclear energy acceptance the techno-economical analysis of different methods for high level waste storage and disposal has to be performed. The aim of this paper is to present technical and economical characteristics of different HLW storage and disposal technologies. The final choice of a particular HLW management method is closely connected to the selection of a fuel cycle type: open or closed. Wet and dry temporary storage has been analyzed including different types of spent fuel pool capacity increase methods, different pool location (at reactor site and away from reactor site) as well as casks and vault system of dry storage. Since deep geological deposition is the only disposal method with a realistic potential, we focused our attention on that disposal technology. Special attention has been given to the new idea of international and regional disposal location. The analysis showed that a coexistence of different storage methods and deep geological deposition is expected in the future, regardless of the fuel cycle type. (author)

Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

International Nuclear Information System (INIS)

Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

2002-01-01

Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

Dry Cask Storage Inspection and Monitoring. Interim Report.

Energy Technology Data Exchange (ETDEWEB)

Bakhtiari, Susan [Argonne National Lab. (ANL), Argonne, IL (United States); Elmer, Thomas W. [Argonne National Lab. (ANL), Argonne, IL (United States); Koehl, Eugene R. [Argonne National Lab. (ANL), Argonne, IL (United States); Wang, Ke [Argonne National Lab. (ANL), Argonne, IL (United States); Raptis, Apostolos C. [Argonne National Lab. (ANL), Argonne, IL (United States); Kunerth, Dennis C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Birk, Sandra M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-03-04

Recently, the U.S. Nuclear Regulatory Commission (NRC) issued the guidance on the aging management of dry storage facilities that indicates the necessity to monitor the conditions of dry cask storage systems (DCSSs) over extended periods of time.1 Part of the justification of the aging management plans is the requirement for inspection and monitoring to verify whether continued monitoring, inspection or mitigation are necessary. To meet this challenge Argonne National Laboratory (ANL) in collaboration with Idaho National Laboratory (INL) is conducting scoping studies on current and emerging nondestructive evaluation/examination (NDE) and online monitoring (OLM) technologies for DCSS integrity assessments. The scope of work plan includes identification and verification of technologies for long-term online monitoring of DCSSs’ crucial physical parameters such as temperature, pressure, leakage and structural integrity in general. Modifications have been made to the current technologies to accommodate field inspections and monitoring. A summary of the scoping studies and experimental efforts conducted to date as well as plans for future activities is provided below.

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

Science.gov (United States)

2012-04-25

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

Status of work on the final repository concept concerning direct disposal of spent fuel rods in fuel rod casks (BSK)

International Nuclear Information System (INIS)

Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.

2008-01-01

The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.

The impact of using reduced capacity baskets on cask fleet size and cask fleet mix

International Nuclear Information System (INIS)

Joy, D.S.; Johnson, P.E.; Andress, D.A.

1993-01-01

The Civilian Radioactive Waste Management System transportation system will encounter a wide range of spent fuel characteristics. Since the Initiative I casks are being designed to transport 10-year-old fuel with a burnup of 35,000 MWd/MTU, there is a good likelihood that a number of the cask shipments will need to be derated in order to meet the Nuclear Regulatory Commission radiation guidelines. This report discusses the impact of cask derating by using reduced-capacity baskets. Cask derating, while enhancing the ability to move spent fuel with a wider range of age and burnup characteristics, increases the number of shipments; the amount of equipment (cask bodies, baskets, etc.); and the number of visits to both shipping and receiving sites required to transport a specific amount of spent fuel

The impact of using reduced-capacity baskets on cask fleet size and cask fleet mix

International Nuclear Information System (INIS)

Joy, D.S.; Johnson, P.E.; Andress, D.A.

1993-01-01

The Civilian Radioactive Waste Management System transportation system will encounter a wide range of spent fuel characteristics. Since the Initiative I casks are being designed to transport 10-year-old fuel with a burnup of 35,000 MWd/MTU, there is a good likelihood that a number of the cask shipments will need to be derated in order to meet the Nuclear Regulatory Commission radiation guidelines. This report discusses the impact of cask derating by using reduced-capacity baskets. Cask derating, while enhancing the ability to move spent fuel with a wider range of age and burnup characteristics, increases the number of shipments; the amount of equipment (cask bodies, baskets, etc.); and the number of visits to both shipping and receiving sites required to transport a specific amount of spent fuel

FACSIM/MRS [Monitored Retrievable Storage]-2: Storage and shipping model documentation and user's guide

International Nuclear Information System (INIS)

Huber, H.D.; Chockie, A.D.; Hostick, C.J.; Otis, P.T.; Sovers, R.A.

1987-06-01

The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for FACSIM/MRS-2, which is also referred to as the back-end model. The FACSIM/MRS-2 model simulates the MRS storage and shipping operations, which include handling canistered spent fuel and secondary waste in the shielded canyon cells, in onsite yard storage, and in repository shipping cask loading areas

WASTES II: Waste System Transportation and Economic Simulation. Version II. User's guide

International Nuclear Information System (INIS)

Shay, M.R.; Buxbaum, M.E.

1986-02-01

The WASTES II model was developed to provide detailed analyses beyond the capabilities of other available models. WASTES uses discrete event simulation techniques to model the generation of commercial spent nuclear fuel, the buildup of spent fuel inventories within the system, and the transportation requirements for the movement of radioactive waste throughout the system. The model is written in FORTRAN 77 as an extension to the SLAM commercial simulation language package. In addition to the pool storage and dry storage located at the reactors, the WASTES model provides a choice of up to ten other storage facilities of four different types. The simulation performed by WASTES may be controlled by a combination of source- and/or destination-controlled transfers that are requested by the code user. The user supplies shipping cask characteristics for truck or rail shipment casks. As part of the facility description, the user specifies which casks the facility can use. Shipments within the system can be user specified to occur optimally, or proximally. Optimized shipping can be used when exactly two destination facilities of the same facility type are open for receipt of fuel. Optimized shipping selects source/destination pairs so that the total shipping distance or total shipping costs in a given year are minimized when both facilities are fully utilized. Proximity shipping sequentially fills the closest facility to the source according to the shipment priorities without regard for the total annual shipments. This results in sub-optimal routing of waste material but can be used to approximate an optimal shipping strategy when more than two facilities of the same type are available to receive waste. WASTES is currently able to analyze each of the commercial spent fuel logistics scenarios specified in the 1985 DOE Mission Plan

Size and transportation capabilities of the existing US cask fleet

International Nuclear Information System (INIS)

Danese, F.L.; Johnson, P.E.; Joy, D.S.

1990-01-01

This study investigates the current spent nuclear fuel cask fleet capability in the United States. In addition, it assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade

Feasibility study for a transportation operations system cask maintenance facility

Energy Technology Data Exchange (ETDEWEB)

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

1991-01-01

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

Feasibility study for a transportation operations system cask maintenance facility

International Nuclear Information System (INIS)

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

1991-01-01

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

Test Plan for Cask Identification Detector

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-29

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

German Approach for the Transport of Spent Fuel Packages after Interim Storage

International Nuclear Information System (INIS)

Wille, Frank; Wolff, Dietmar; Droste, Bernhard; Voelzke, Holger

2014-01-01

In Germany the concept of dry interim storage of spent nuclear fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently (in 2012) revised 'Guidelines for dry cask storage of spent nuclear fuel and heat-generating waste' by the German Waste management Commission (ESK) which are very similar to the former RSK (reactor safety commission) guidelines. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric seals under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts are issues as well. This paper describes the state

Thick nickel plating of spent fuel transport and storage casks CASTOR and POLLUX

International Nuclear Information System (INIS)

Wilbuer, K.

1991-01-01

Spent fuel elements have to be safely handled in containers for transport and storage. These large casks (100-120 t) are made by various firms according to the specifications given by the nuclear plant operator. For shielding and protection of the hazardous material, the casks' inner surface is coated with a nickel plating about 3000 μm thick. The product and the production process are subject to very stringent requirements, due to the hazardous potential of the material to be shipped or stored. Therefore, both the extremely high quality standards to be met by the nickel plating and the dimensions and capability of the plating plant required for the process are problems that cannot be solved by a usual commercial plating plant. The new concept and process that had to be established are explained in the paper. (orig./MM) [de

Verification of Vitrified High-Activity Waste Stored in a CASTOR HAW 20/28 CG Cask by Simulated Baseline Comparison

International Nuclear Information System (INIS)

Shephard, A.; Arenas-Carrasco, J.; Dratschmidt, H.; De-Baere, P.; Af Ekenstam, G.; Lebrun, A.

2010-01-01

The verification process for the vitrification of high-activity waste (HAW) focuses on maintaining the continuity-of-knowledge of special nuclear material (SNM) as it traverses a vitrification facility. However, the inaccessible nature of a vitrification facility presents an obstacle to the deployment of conventional safeguards, albeit the process area of a vitrification facility is effectively a hot cell. The employment of remotely operated NDA hardware/DA sample equipment inside the process area would be problematic-at-best and the alternative of continuous monitoring would draw heavily on the critical resource of inspector time. In response to the aforementioned constraints, the IAEA and Euratom opted to develop a new method which focuses on the verification of SNM after the vitrified HAW has been sealed in storage casks. The new method verifies the presence of the vitrified HAW through the comparison of total neutron count rates collected at points around a cask with those predicted by Monte Carlo simulation. The model includes a dual N50 neutron slab detector (custom design by Euratom) and a CASTOR HAW 20/28 CG storage cask configured with the operator declared contents. By comparison of the simulated neutron emission pattern and field measurements, the displacement of Pu and U is evident from a detectable neutron signal defect. Because the spontaneous fission of 244 Cm is the dominant neutron source in vitrified HAW, the 244 Cm/Pu and 244 Cm/U mass ratios must be known in order to relate the neutron signal outside the cask to the amounts of Pu and U stored inside. These mass ratios can be determined from HAW samples collected by the inspectorates from the accountability tanks and analyzed by DA. The absence of separation of SNM from the HAW is verified by other measures. To ensure the validity of the simulation, sources of uncertainty were systematically addressed and quantified. This new verification method effectively removes the need for NDA equipment

The century storage facility. Why we have to give thoughts to the castor halls; Die Jahrhundert-Lager. Warum wir uns dringend Gedanken ueber die Castor-Hallen machen muessen

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-01-15

The authors discuss the situation of the German interim storage facilities for high-level nuclear waste. The Castor casks are supposed to be licensed as storage casks for 40 years. It is already foreseeable that after 40 years a final repository will not be available. This means that the Castor casks and the interim storage halls will have to be operated at least for the 21st century without valid licensing.

Thermal Test in accordance with Mesh Size at Opening of the Inlets and Outlets of Concrete Storage Cask

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Yu, S. H.; Lee, J. C.; Choi, W. S. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The concrete storage cask must be designed to have heat removal capabilities with appropriate reliability. However, the thermal conductivity of concrete is not adequate for this purpose. The American Concrete Institute standard ACI-349 specifies a limit of 66 .deg. C for the normal operating temperature of concrete, except for the local areas, which may not exceed 93 .deg. C, and a short-term or accident temperature limit of no more than 177 .deg. C. Therefore, a passive heat removal system was designed to maintain the temperatures of the fuel-assembly cladding material and concrete storage cask components within these allowable limits. The passive heat-removal system consists of four inlets and four outlets, and their openings are covered by screens of mesh structure to prevent debris or wildlife from entering the ventilation ducts. Depending on its mesh size, each screen will have a different effect on the heat removal of the concrete storage cask. This paper discusses the experimental approach used in the present study to evaluate the heat removal performance under normal conditions in accordance with the mesh size of the screen installed at the opening of the inlets and outlets. The main results of the study are described below. The mesh size of the screen had an insignificant effect on the temperature rise of the canister surface and the over-pack surface.

Safety aspects of dry spent fuel storage and spent fuel management

International Nuclear Information System (INIS)

Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)

Size and transportation capabilities of the existing U.S. cask fleet

International Nuclear Information System (INIS)

Danese, F.L.; Johnson, P.E.; Joy, D.S.

1990-01-01

This paper investigates the current spent nuclear fuel cask fleet capability in the United States. It assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade

Spent fuel shipping cask development status

International Nuclear Information System (INIS)

Henry, K.H.; Lattin, W.C.

1989-01-01

The Nuclear Waste Policy Act of 1982 (NWPA) authorized the US Department of Energy (DOE) to establish a national system for the disposal of spent nuclear fuel and high-level radioactive waste from commercial power generation, and established the Office of Civilian Radioactive Waste Management (OCRWM) within the DOE-Headquarters (DOE-HQ) to carry out these duties. A 1985 presidential decision added the disposal of high-level radioactive waste generated by defense programs to the national disposal system. A primary element of the disposal program is the development and operation of a transportation system to move the waste from its present locations to the facilities that will be included in the waste management system. The primary type of disposal facility to be established is a geologic repository; a Monitored Retrievable Storage (MRS) facility may also be included as an intermediate step in the nuclear waste disposal process. This paper focuses on the progress and status of one facet of the transportation program--the development of a family of shipping casks for transporting spent fuel from nuclear power reactor sites to the repository of MRS facility

FACSIM/MRS (Monitored Retrievable Storage)-2: Storage and shipping model documentation and user's guide

Energy Technology Data Exchange (ETDEWEB)

Huber, H.D.; Chockie, A.D.; Hostick, C.J.; Otis, P.T.; Sovers, R.A.

1987-06-01

The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for FACSIM/MRS-2, which is also referred to as the back-end model. The FACSIM/MRS-2 model simulates the MRS storage and shipping operations, which include handling canistered spent fuel and secondary waste in the shielded canyon cells, in onsite yard storage, and in repository shipping cask loading areas.

TMI abnormal wastes disposal options

International Nuclear Information System (INIS)

Ayers, A.L. Jr.

1984-03-01

A substantial quantity of high beta-gamma/high-TRU contaminated wastes are expected from cleanup activities of Unit 2 of the Three Mile Island Nuclear Power Station. Those wastes are not disposable because of present regulatory constraints. Therefore, they must be stored temporarily. This paper discusses three options for storage of those wastes at the Idaho National Engineering Laboratory: (1) storage in temporary storage casks; (2) underground storage in vaults; and (3) storage in silos at a hot shop. Each option is analyzed and evaluated. Also included is a discussion of future disposal strategies, which might be pursued when a suitable federal or commercial repository is built

Comparison of elastic and inelastic analysis and test results for the defense high level waste shipping cask

International Nuclear Information System (INIS)

Zimmer, A.; Koploy, M.A.; Madsen, M.M.

1991-01-01

In the early 1980s, the US DOE/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as fully complies with all applicable DOE, Nuclear Regulatory Commission, and DOT regulations. General Atomics designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Labs. (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This paper will compare the results of the two analytical approaches and with model testing results. The purpose of this work is to provide data to support licensing of the DHLW cask and to support the acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

Transportation capabilities of the existing cask fleet

International Nuclear Information System (INIS)

Johnson, P.E.; Wankerl, M.W.; Joy, D.S.

1991-01-01

This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the DOE consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated

Transportation capabilities of the existing cask fleet

International Nuclear Information System (INIS)

Johnson, P.E.; Joy, D.S.; Wankerl, M.W.

1991-01-01

This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 46 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the Department of Energy (DOE) consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated. 5 refs., 4 tabs

Seal performance of thermal aged metal gasket of dual purpose metal cask for interim spent fuel storage after external impact load

International Nuclear Information System (INIS)

Takeshi Yokoyama; Masami Kato; Satoshi Itooka

2005-01-01

As for interim storage for spent nuclear fuels using dual purpose dry metal cask in Japan, we recognize one of the important technical issues that there is a possibility for the cask with degraded metal gasket during storage to apply to transportation. In our study until 2003 focused on degradation of important components for the cask safety performance during storage and application to transportation, for metal gasket, we conducted property tests for degradation and influence of lid movement on seal performance, and furthermore verification tests. From the results, we developed the method to evaluate leak rate from lid with degraded metal gasket at accidents during transportation and in addition, we found as follows: Lid would hardly move and leak rate would not increase seriously during fire event. Leak rate from lid with degraded metal gasket could be evaluated by using results of leak rate trend depending on horizontal displacement of lid by external impact load. So, with regard to influence of lid movement on seal performance, we conducted additional test for extending horizontal displacement in lid moving in 2004. In addition, seal performance was discussed from the results, both previous and latest test. (authors)

Interfacing the existing cask fleet with the MRS

International Nuclear Information System (INIS)

Doman, J.W.; Hahn, R.E.

1992-01-01

This paper reports that the Department of Energy (DOE) is considering the possibility of using the existing fleet of casks to achieve spent fuel receipt at the Monitored Retrievable Storage (MRS) facility. The existing cask fleet includes the NLI-1/2, the NAC-LWT, the TN-8 (and TN-8L), the TN-9, and the IF-300 casks. Other casks may be available, but their status is not certain. Use of the existing cask fleet at the MRS places additional design requirements on the system, and specifically affects the cask-to-MRS interface. The decision to use the existing cask fleet also places additional demands on training needs and operator certification, and the configuration management system. Some existing cask designs may not be able to mate with a bottom opening hot cell MRS. Use of the existing cask fleet also greatly increases the number of shipments that must be received, to the point that a facility larger than originally envisioned may be required

Impact of more conservative cask designs of the CRWMS transportation system

International Nuclear Information System (INIS)

Joy, D.S.; Pope, R.B.; Johnson, P.E.

1993-01-01

The Office of Civilian Radioactive Waste Management has been working since the mid-1980s to develop a cask fleet, which will include legal weight truck and rail/barge casks for the transport of spent nuclear fuel (SNF) from reactors to Civilian Radioactive Waste Management System SNF receiving sites. The cask designs resulting from this effort have been identified as Initiative I casks. In order to maximize payloads, advanced technologies have been incorporated in the Initiative I cask designs, and some design margins have been reduced. Due to the wide range of the characteristics (age/burnup) of the spent fuel assemblies to be transported in the Initiative I casks, it has become apparent that a significant portion of the shipments of the Initiative I casks could not be loaded to their design capacity. Application of a more conventional cask design philosophy might result in new generation casks that would be easier to license, have more operational flexibility as to the range of age/burnup fuel that could be transported at full load, and be easier to fabricate. In general, these casks would have a lower capacity than the currently proposed Initiative I casks, thereby increasing the transportation impacts and the transportation costs

Development of integrated cask body and base plate

International Nuclear Information System (INIS)

Sasaki, T.; Koyama, Y.; Yoshida, T.; Wada, T.

2015-01-01

The average of occupancy of stored spent-fuel in the nuclear power plants have reached 70 percent and it is anticipated that the demand of metal casks for the storage and transportation of spent-fuel rise after resuming the operations. The main part of metal cask consists of main body, neutron shield and external cylinder. We have developed the manufacturing technology of Integrated Cask Body and Base Plate by integrating Cask Body and Base Plate as monolithic forging with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Here, we report the manufacturing technology, code compliance and obtained properties of Integrated Cask body and Base Plate. (author)

Storage fee analysis for a retrievable surface storage facility

International Nuclear Information System (INIS)

Field, B.B.; Rosnick, C.K.

1973-12-01

Conceptual design studies are in progress for a Water Basin Concept (WBC) and an alternative Sealed Storage Cask Concept (SSCC) of a Retrievable Surface Storage Facility (RSSF) intended as a Federal government facility for storing high-level radioactive wastes until a permanent disposal method is established. The RSSF will be a man-made facility with a design life of at least 100 y, and will have capacity to store all of the high-level waste from the reprocessing of nuclear power plant spent fuels generated by the industry through the year 2000. This report is a basic version of ARH-2746, ''Retrievable Surface Storage Facility, Water Basin Concept, User Charge Analysis.'' It is concerned with the issue of establishing a fee to cover the cost of storing nuclear wastes both in the RSSF and at the subsequent disposal facility. (U.S.)

Decree of the Czechoslovak Atomic Energy Commission No. 8 as of 25 June 1981 on the testing of equipment for radioactive waste transport, storage and disposal

International Nuclear Information System (INIS)

1995-01-01

The Decree stipulates that manufacturers and users of equipment for radioactive waste transportation, storage and disposal are obliged to have the equipment tested. This duty concerns radioactive waste transport casks, shielding containers, etc., except for nuclear fuel transporting facilities. Authorization to act as the national testing body was granted to the Institute for Research, Production and Application of Radioisotopes. The Decree entered into force on 1 July 1981. (J.B.)

Transportation of high-level waste and spent fuel

International Nuclear Information System (INIS)

Carlson, J.H.; Lake, W.H.; Thompson, J.H.

1993-01-01

The Office of Civilian Radioactive Waste Management (OCRWM) transportation program is a multifaceted undertaking to transport spent nuclear fuel from commercial reactors to temporary and permanent storage facilities commencing in 1998. One of the significant ingredients necessary to achieving this goal is the development and acquisition of shipping casks. Efforts to design and acquire high capacity casks is ongoing, as are efforts to purchase casks that can be made available using current technology. By designing casks that are optimized to the specifications of the older cooler spent fuel that will be shipped, and by designing to current NRC requirements, OCRWM's new generation of spent fuel casks will be more efficient and at least as safe as current cask designs. (J.P.N.)

NAC-1 cask dose rate calculations for LWR spent fuel

International Nuclear Information System (INIS)

CARLSON, A.B.

1999-01-01

A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation

Re-evaluation of monitored retrievable storage concepts

International Nuclear Information System (INIS)

Fletcher, J.F.; Smith, R.I.

1989-04-01

In 1983, as a prelude to the monitored retrievable storage (MRS) facility conceptual design, the Pacific Northwest Laboratory (PNL) conducted an evaluation for the US Department of Energy (DOE) that examined alternative concepts for storing spent LWR fuel and high- level wastes from fuel reprocessing. The evaluation was made considering nine concepts for dry away-from-reactor storage. The nine concepts evaluated were: concrete storage cask, tunnel drywell, concrete cask-in-trench, open-cycle vault, metal casks (transportable and stationary), closed-cycle vault, field drywell, and tunnel-rack vault. The purpose and scope of the re-evaluation did not require a repetition of the expert-based examinations used earlier. Instead, it was based on more detailed technical review by a small group, focusing on changes that had occurred since the initial evaluation was made. Two additional storage concepts--the water pool and the horizontal modular storage vault (NUHOMS system)--were ranked along with the original nine. The original nine concepts and the added two conceptual designs were modified as appropriate for a scenario with storage capacity for 15,000 MTU of spent fuel. Costs, area requirements, and technical and historical data pertaining to MRS storage were updated for each concept

Re-evaluation of monitored retrievable storage concepts

Energy Technology Data Exchange (ETDEWEB)

Fletcher, J.F.; Smith, R.I.

1989-04-01

In 1983, as a prelude to the monitored retrievable storage (MRS) facility conceptual design, the Pacific Northwest Laboratory (PNL) conducted an evaluation for the US Department of Energy (DOE) that examined alternative concepts for storing spent LWR fuel and high- level wastes from fuel reprocessing. The evaluation was made considering nine concepts for dry away-from-reactor storage. The nine concepts evaluated were: concrete storage cask, tunnel drywell, concrete cask-in-trench, open-cycle vault, metal casks (transportable and stationary), closed-cycle vault, field drywell, and tunnel-rack vault. The purpose and scope of the re-evaluation did not require a repetition of the expert-based examinations used earlier. Instead, it was based on more detailed technical review by a small group, focusing on changes that had occurred since the initial evaluation was made. Two additional storage concepts--the water pool and the horizontal modular storage vault (NUHOMS system)--were ranked along with the original nine. The original nine concepts and the added two conceptual designs were modified as appropriate for a scenario with storage capacity for 15,000 MTU of spent fuel. Costs, area requirements, and technical and historical data pertaining to MRS storage were updated for each concept.

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

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Geological and seismological characteristics for... § 72.103 Geological and seismological characteristics for applications for dry cask modes of storage on... foundation and geological investigation, literature review, and regional geological reconnaissance show no...

SGN multipurpose dry storage technology applied to the Italian situation

International Nuclear Information System (INIS)

Giorgio, M.; Lanza, R.

1999-01-01

SGN has gained considerable experience in the design and construction of interim storage facilities for spent fuel and various nuclear waste, and can therefore propose single product and multipurpose facilities capable of accommodating all types of waste in a single structure. The pooling of certain functions (transport cask reception, radiation protection) and the choice of optimized technologies to answer the specific needs of clients (transfer of nuclear packages by shielded handling cask or nuclearized crane), the use of the same type of storage pit to cool the heat releasing packages (vitrified nuclear waste, fuel elements) makes it possible to propose industrially proven and cost-effective solutions. Studies carried out for the Dutch company COVRA (HABOG facility currently under implementation phase) provide an example of a multipurpose dry storage facility designed to store spent fuel, vitrified reprocessing waste, cemented hulls and end-pieces, cemented technological waste and bituminized waste from fuel reprocessing, i e. high level waste and intermediate level wastes. The study conducted by SGN and GENESI (an Italian consortium formed by Ansaldo's Nuclear Division and Fiat Avio), on behalf of the Italian utility ENEL, offers another example of the multipurpose dry storage facility designed to store in a centralised site all the remaining irradiated fuel elements plus the vitrified waste. This paper presents SGN's experience through a short description of reference storage facilities for various types of products (HLW and spent fuel). It continues with the typical application to the Italian situation to show how these proven technologies are combined to obtain multipurpose facilities tailored to the client's specific requirements. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

Barkauskas, V.; Plukiene, R.; Plukis, A.

2016-01-01

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

Thermal Analysis of Concrete Storage Cask with Bird Screen Meshes

Energy Technology Data Exchange (ETDEWEB)

Lee, Ju-Chan; Bang, K.S.; Yu, S.H.; Cho, S.S.; Choi, W.S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, a thermal analysis of the cask with bird screen meshes has been performed using a porous media model. The overpack consists of a structural material, a concrete shielding, and a ventilation system. Heat is removed from the cask to the environment by a passive means only. Air inlet and outlet ducts are installed at the bottom and top of the cask for a ventilation system. Bird screen meshes are installed at the air inlet and outlet ducts to inhibit intrusion of debris from the external environment. The presence of this screens introduce an additional resistance to air flow through the ducts. Five types of meshes for bird screen were considered in this study. The bird screen meshes at the inlet and outlet vents reduce the open area for flow by about 44 - 79 %. Flow resistance coefficients for porous media model were deduced from the fluid flow analysis of bird screen meshes. Thermal analyses for the concrete cask have been carried out using a porous media model. The analysis results agreed well with the test results. Therefore, it was shown that the porous media model for the screen mesh was established to estimate the cask temperatures.

Thermal Analysis of Concrete Storage Cask with Bird Screen Meshes

International Nuclear Information System (INIS)

Lee, Ju-Chan; Bang, K.S.; Yu, S.H.; Cho, S.S.; Choi, W.S.

2016-01-01

In this study, a thermal analysis of the cask with bird screen meshes has been performed using a porous media model. The overpack consists of a structural material, a concrete shielding, and a ventilation system. Heat is removed from the cask to the environment by a passive means only. Air inlet and outlet ducts are installed at the bottom and top of the cask for a ventilation system. Bird screen meshes are installed at the air inlet and outlet ducts to inhibit intrusion of debris from the external environment. The presence of this screens introduce an additional resistance to air flow through the ducts. Five types of meshes for bird screen were considered in this study. The bird screen meshes at the inlet and outlet vents reduce the open area for flow by about 44 - 79 %. Flow resistance coefficients for porous media model were deduced from the fluid flow analysis of bird screen meshes. Thermal analyses for the concrete cask have been carried out using a porous media model. The analysis results agreed well with the test results. Therefore, it was shown that the porous media model for the screen mesh was established to estimate the cask temperatures

The NINO [No Inspector, No Operator system] cask-loading safeguards system

International Nuclear Information System (INIS)

Fiarman, S.

1987-01-01

It is, in general difficult to determine by means of camera-surveillance techniques what is loaded into spent-fuel casks being prepared for shipment from light-water reactors to other reactors, reprocessing facilities, or long-term storage. Furthermore, the expected high frequency of cask loadings in the coming years would place too great a burden on the IAEA and Euratom inspectorates if each had to be observed by an inspector. For the case of shipment to other reactors and reprocessing facilities, the casks are soon opened and, in principle, their contents could be ascertained by direct inspection. In the case of long-term-storage facilities, the casks would stay sealed for years, thereby requiring the IAEA to know positively how many spent-fuel assemblies were loaded at the reactor and to have a continuity of knowledge of the cask's contents. It has been proposed instead that the facility operator place the cask seal on the cask within the field of view of a surveillance system linked to the cask seal. This solution, however, may not provide enough credibility for acceptance by the safeguards community. This paper presents an alternative to both inspector presence at cask loading and operator assistance in applying seals; this alternative is called the No Inspector, No Operator system (NINO)

Influence of local regulations on TN dual purpose BWR casks

International Nuclear Information System (INIS)

Samson, P.; Neider, T.

1999-01-01

Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

Storage container for radioactive fuel elements

International Nuclear Information System (INIS)

1984-01-01

The interim storage cask for spent fuel elements or the glass moulds for high-level radioactive waste are made up of heat-resistant, reinforced concrete with chambers and highgrade steel lining. Cooling systems with natural air circulation are connected with the chambers. (HP) [de

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)

Cask ownership: Options and strategic factors

International Nuclear Information System (INIS)

Smith, C.W.

1986-01-01

Because of the potential number of casks available through utility modular storage programs, it is imperative that the planning for the provision and operation of casks under the NWPA program include consideration of the utility owned casks. As to the remainder of the cask requirements for implementation of the NWPA, the author believes that the cost factor is an artificial one for determining the benefits to the taxpayers and ratepayers for cask ownership and that the decision should be made on the basis of capability of the industry to perform on a competitive bid basis and assurance that the shipments will be made on a timely, safe and cost effective basis. If the procurement process is structured to rally permit competitive bidding on spent fuel shipping services, the competition in the market place will assure that DOE and the ratepayers, receive safe, high quality, and cost effective transportation proposals from very capable companies

Utility oversight of Cask System Development Program

International Nuclear Information System (INIS)

Vincent, J.A.; Jordan, J.M.; Schwartz, M.H.

1993-01-01

This paper will present the electric utility industry's perspective on the status and scope of the DOE's Office of Civilian Radioactive Waste Management's (DOE/OCRWM) transportation cask systems development activities, including the Cask Systems Development Program (CSDP) Initiative I transportation cask projects. This presentation is particularly timely because the CSDP Independent Management Review Group (IMRG), os which one of the authors is a member, completed an objective assessment of OCRWM's transportation cask system development activities and issued its first report in late August 1992. The perspective on these cask systems development activities that will be presented reflects conclusions based on (1) the industry's review of CSDP Preliminary and Draft Final Design Reports for the Initiative I cask projects, (2) the activities of one of the authors as a member of the IMRG, and (3) the positions that the industry has consistently taken on what it believes to be the appropriate scope and pace of the CSDP and its integration with other OCRWM activities. Background information on the OCRWM transportation cask systems development activities and the relevant industry activities will also be provided

Analysis of DCI cask drop test onto reinforced concrete pad

International Nuclear Information System (INIS)

Ito, C.; Kato, Y.; Hattori, S.; Shirai, K.; Misumi, M.; Ozaki, S.

1993-01-01

In a cask-storage facility, a cask may be subjected to an impact load as a result of a free drop onto the floor because of cask mishandling. We performed drop tests of casks onto a reinforced concrete (RC) slab representing the floor of a facility as well as simulation analysis [Kato et al]. This paper describes the details of the FEM analysis and calculated results and compares them with the drop test results. (J.P.N.)

Used Fuel Cask Identification through Neutron Profile

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-20

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

Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

International Nuclear Information System (INIS)

Leber, A.; Graf, W.; Hueggenberg, R.

2004-01-01

With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT

BR-100 spent fuel shipping cask development

International Nuclear Information System (INIS)

McGuinn, E.J.; Childress, P.C.

1990-01-01

Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B ampersand W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs

Method to mount defect fuel elements i transport casks

International Nuclear Information System (INIS)

Borgers, H.; Deleryd, R.

1996-01-01

Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

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

Energy Technology Data Exchange (ETDEWEB)

McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

2018-01-01

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

Return transport of processed radioactive waste from France and Great Britain

International Nuclear Information System (INIS)

2010-11-01

The report on returning transport and interim storage of processed radioactive waste from France and Great Britain in vitrified block containers covers the following issues: German contracts with radioactive waste processing plants concerning the return of processed waste to Germany; optimized radioactive waste processing using vitrified block containers; the transport casks as basic safety with respect to radiation protection; interim storage of processes high-level waste by GNS in Gorleben; licensing, inspections and declarations; quality assurance and control.

Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

International Nuclear Information System (INIS)

Carbajo, J.J.; Lindner, C.N.

1992-01-01

The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car

Application of constraint satisfaction algorithms for conditioning and packing activated control rod assemblies in MOSAIK"R-casks. Application of constraint satisfaction algorithms for conditioning and packaging 160 control rod assemblies

International Nuclear Information System (INIS)

Harding, P.J.

2017-01-01

In the wake of the decommissioning of numerous nuclear power plants in Germany, techniques to reduce the number of costly waste casks or containers are sought after. The large bandwidth of limits (dose rate, mass, individual nuclide activities, chemical composition...) the waste packages have to comply with for both interim storage facilities and the repository Konrad render the manual planning of packaging concepts prohibitive. However, in the past, the planning for packaging has been performed in this way, albeit on the basis of several facilitating assumptions. Surprisingly, to the best of our knowledge, the automated computer-assisted generation of packaging plans for radioactive waste has not been demonstrated previously. In this talk we demonstrate how the conditioning and packing of 160 control rod assemblies was optimised using constraint satisfaction algorithms. These algorithms can be executed by a computer in a few minutes, thus considerably accelerating the generation of packaging plans, while optimising the utilisation of the waste casks and containers with respect to mass, activity, dose rate, etc. This automated and computer-assisted procedure took into account complex logistical boundary conditions present during decommissioning, such as space requirements, the sequence of the waste and the (lack of) availability of suitable waste casks. In addition, packaging concepts based on several scenarios (cask availability, space requirements,...) were easily and automatically generated once the packaging rules had been coded. We demonstrate the successful application of these algorithms to a real packaging campaign of control rod assemblies of a boiling water reactor, for which excellent results were achieved. We also present an outlook of a much larger scale project, in which the logistics and storage of radioactive waste packages is mathematically optimised. Finally, we give prospects on these techniques to others, similar logistical problems currently

Radioactive fuel cask railcar humping study

International Nuclear Information System (INIS)

Wilson, L.T.

1978-01-01

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

COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

International Nuclear Information System (INIS)

Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

2003-01-01

Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal

Update of Nuclear Waste Policy Act transportation activities

International Nuclear Information System (INIS)

Callaghan, E.F.

1987-01-01

As directed by the Nuclear Waste Policy Act of 1982 (NWPA), the Department of Energy (DOE) is developing a nationwide system for transporting spent nuclear fuel and high-level radioactive waste from commercial power plants to deep geologic repositories for disposal. Plans for the transportation system will consider the following factors: the President's 1985 decision to co-locate some defense high-level waste with commercial waste in a repository, the NWPA requirement that the private sector be used to the fullest extent possible in developing and operating the system, and the possible approval by Congress of the DOE's proposal for a Monitored Retrievable Storage (MRS) facility, submitted in March 1987. (The MRS, if approved, would provide for the consolidation, packaging, and perhaps the temporary storage of spent fuel from reactors.) The ''Transportation Business Plan'', published in January 1986, reflects these considerations. The transportation system, when operational, will consist of two elements: (1) the cask system, which includes the transportation casks, the vehicular conveyances, tie-downs, and associated equipment for handling the casks; and (2) the transportation support system which is comprised of facilities, equipment, and services to support waste transportation. Development of the transportation system incorporates the following work elements: operational planning, support systems development, cash system development, systems analysis, and institutional activities. This paper focusses on the technical aspects of the system

Test Plan for the Boiling Water Reactor Dry Cask Simulator

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-01

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

Dry interim storage of radioactive material in Germany

International Nuclear Information System (INIS)

Drobniewski, Christian; Palmes, Julia

2013-01-01

In accordance with the waste management concept in Germany, spent fuel is stored in interim storage facilities for a period of up to 40 years until deposition in a geological repository. In twelve on-site interim storages in the vicinity or directly on the sites of the nuclear power plants, spent fuel elements from reactor operation are stored after the necessary period of decay in wet storage basins inside the reactors. Additionally, three central interim storage facilities for storage of spent fuel of different origin are in operation. The German facilities realize the concept of dry interim storage in metallic transport and storage casks. The confinement of the radioactive material is ensured by the double lid system of the casks, of which the leak tightness is monitored constantly. The casks are constructed to provide adequate heat removal and shielding of gamma and neutron radiation. Usually the storage facilities are halls of thick concrete structures, which ensure the removal of the decay heat by natural convection. The main safety goal of the storage concept is to prevent unnecessary exposure of persons, material goods and environment to ionizing radiation. Moreover any exposure should be kept as low as reasonable achievable. To reach this goal the containment of the radioactive materials, the disposal of decay heat, the sub criticality and the shielding of ionizing radiation has to be demonstrated by the applicant and verified by the licensing authority. In particular accidents, incidents and disasters have to be considered in the facility and cask design. This includes mechanical impacts onto the cask, internal and external fire, and environmental effects like wind, rain, snowfall, flood, earthquakes and landslides. In addition civilizatoric influences like plane crashes and explosions have to be taken into account. In all mentioned cases the secure confinement of the radioactive materials has to be ensured. On-site storage facilities have to consider the

Dry storage technologies: Optimized solutions for spent fuels and vitrified residues

International Nuclear Information System (INIS)

Roland, Vincent; Verdier, Antoine; Sicard, Damien; Neider, Tara

2006-01-01

In many countries, fuel cycle and waste policies influence the way operators organize waste management. These policies help drive progress and improvements in areas such as waste minimization programs, conditioning or industrial transformation before final or intermediate conditioning. The criteria that lead to different choices include economic factors, the presence or absence of a wide range of options such as transport, and reprocessing and recycling policies. The current international trend towards expanding Spent Fuel Interim Dry Storage capabilities goes with an improvement of the performance of the proposed systems which have to accommodate Spent fuel Assemblies characterized by ever increasing burn-up, fissile isotopes contents, thermal releases, and total inventory. Due to heterogeneous worldwide reactor pools and specific local constraints the proposed solutions have also to cope with a wide variety of fuel design. The Spent Fuel Assemblies stored temporarily for cooling may have to be transported either to reprocessing facilities or to interim storage facilities before direct disposal; it is the reason why the retrievability, including or not the need of transportation of the proposed systems, is often specified by the utilities for the design of their storage systems and sometimes required by law. In most cases, the producers of spent fuel require a large capacity that is cumulated over many years, each reload at a time. Then the key criterion is maximum modularity. Furthermore, the up front capital costs required for this type of solution has to be attractive for the investor. Two solutions, dual purpose metal casks of the TN TM 24 family or dual purpose or single purpose concrete shielded welded canisters such as NUHOMS R , implemented by COGEMA LOGISTICS, and TRANSNUCLEAR Inc. offer flexibility and modularity and have been adapted also to quite different fuels. Among what influences the choice, we can consider: - In favor of metal casks: Minimal

Development of high performance and low radio activation concrete material for concrete cask

International Nuclear Information System (INIS)

Shirai, Koji; Sonobe, Ryoji

2005-01-01

For the realization of the long-term storage of the nuclear spent fuel with the concrete cask technology, a low radio activation high performance concrete was developed, which contains extremely small quantity of Eu and Co and assures enough heat-resistance and durability for degradation. Firstly, the activation analysis was performed to estimate the allowable content limit of their quantities according to the rules issued by Japanese government for determining the classification of the radioactive waste. Secondly, various candidate materials were sampled and irradiated to find out the activation level. As a result, as the optimum concrete mix, the combination of limestone and white fused alumina aggregates with fry-ash was chosen. Moreover, the basic characteristics of the candidate concrete (workability, strength under high temperature, heat conductivity and so on) were evaluated, and the thermal cracking test was executed with hollow cylinders. Finally, the developed concrete material seems to be suitable for the long-term use of concrete cask considering the low activation, high heat resistance and durability during storage. (author)

Disposal or radioactive waste. Challenges and approaches; Entsorgung von radioaktiven Abfaellen. Herausforderungen und Loesungsansaetze

Energy Technology Data Exchange (ETDEWEB)

Bode, Matthias; Marx, Steffen; Schacht, Gregor [Hannover Univ. (Germany). Inst. fuer Massivbau

2017-01-15

International there's a consensus that the high active waste has to be disposed in deep geological repository. After the reform of the site selection process it's still a long way from having a fully operational final repository in Germany. Currently there is no operational final repository for high active waste around the world. Presently the high active waste is stored regarding the concept of the dry interim storage. Thereby the waste is stored in special cast-iron casks for storage and transport of radioactive material. These casks are located in reinforced concrete buildings, the centralised and decentralised interim storage facilities. This storage is licensed for 40 years. After expiration of the licenses between 2034 and 2047 the high active waste shall disposed in a deep geological final repository in Germany. Due to several delays there won't be an operational available final repository when the licenses will expire. Therefore a new concept for the storage after the expirations has to be developed. Such a ''Ueberbrueckungslagerung'' will be necessary for several decades up to more than a century regarding different prognoses. This article describes and discuss different solutions.

Development of high-strength aluminum alloys for basket in transport and storage cask for high burn-up spent fuel

International Nuclear Information System (INIS)

Maeguchi, T.; Sakaguchi, Y.; Kamiwaki, Y.; Ishii, M.; Yamamoto, T.

2004-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has developed high-strength borated aluminum alloys (high-strength B-Al alloys), suitable for application to baskets in transport and storage casks for high burn-up spent fuels. Aluminum is a suitable base material for the baskets due to its low density and high thermal conductivity. The aluminum basket would reduce weight of the cask, and effectively release heat generated by spent fuels. MHI had already developed borated aluminum alloys (high-toughness B-Al alloy), and registered them as ASME Code Case ''N-673''. However, there has been a strong demand for basket materials with higher strength in the case of MSF (Mitsubishi Spent Fuel) casks for high-burn up spent fuels, since the basket is required to stand up to higher stress at higher temperature. The high-strength basket material enables the design of a compact cask under a limitation of total size and weight. MHI has developed novel high-strength B-Al alloys which meet these requirements, based on a new manufacturing process. The outline of mechanical and metallurgical characteristics of the high-strength B-Al alloys is described in this paper

Management of spent fuel from power and research reactors using CASTOR and CONSTOR casks and licensing experience worldwide

International Nuclear Information System (INIS)

Becher, D.

2003-01-01

An overview of the spent fuel storage in CASTOR and CONSTOR casks during the last 30 years is made. Design characteristics of the both types of casks are presented. CASTOR casks fulfill both the requirements for type B packages according to the IAEA requirements covering different accident situations in storage sites. Analyses of nuclear and thermal behavior and strength are carried out for CONSTOR concept. Special experimental program for verification of mechanical and thermomechanical properties is implemented. Licensing experience of the casks in German storage facilities is presented. Special modifications of CASTOR casks for WWER-440 and RBMK fuel assemblies have been designed for implementation in Eastern Europe. Contracts for GNB spent fuel casks delivery are concluded with Czech Republic, Slovakia, Hungary and Lithuania

Interim storage of radioactive waste packages

International Nuclear Information System (INIS)

1998-01-01

This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

Performances of TN {sup registered} 24 E. An AREVA used fuel transport and interim storage cask for the German market

Energy Technology Data Exchange (ETDEWEB)

Brion, Thomas [AREVA TN International, Montigny Le Bretonneux (France)

2013-07-01

Part of the AREVA Group, TN International offers a complete range of transport and interim storage solutions for radioactive materials throughout the entire nuclear fuel cycle. A world leader in its sector, TN International has supported for 50 years the expansion of the nuclear industry, in particular by providing expertise in secure packing systems for the storage of used fuel assemblies. As an answer to EON and EnBW, two German utilities, needs, TN International has designed and manufactured the TN {sup registered} 24E cask, offering the following high level performances: 1. transport and storage over a period of 40 years of up to 21 PWR spent nuclear fuel (SNF), allowing for example to load up to 17 MOX fuel assemblies and 4 UOX SNF. 2. high flexibility in the fuel assemblies loading plans, inducing no general predefined constraints with regards to the MOX or UOX fuel positions in the basket of the cask Safety margin related to radioprotection, thermal and mechanical behaviour of the fuel assemblies can be calculated loading plan per loading plan. (orig.)

FACSIM/MRS-1: Cask receiving and consolidation model documentation and user's guide

International Nuclear Information System (INIS)

Lotz, T.L.; Shay, M.R.

1987-06-01

The Pacific Northwest Laboratory (PNL) has developed a stochastic computer model, FACSIM/MRS, to assist in assessing the operational performance of the Monitored Retrievable Storage (MRS) waste-handling facility. This report provides the documentation and user's guide for the component FACSIM/MRS-1, which is also referred to as the front-end model. The FACSIM/MRS-1 model simulates the MRS cask-receiving and spent-fuel consolidation activities. The results of the assessment of the operational performance of these activities are contained in a second report, FACSIM/MRS-1: Cask Receiving and Consolidation Performance Assessment (Lotz and Shay 1987). The model of MRS canister storage and shipping operations is presented in FACSIM/MRS-2: Storage and Shipping Model Documentation and User's Guide (Huber et al. 1987). The FACSIM/MRS model uses the commercially available FORTRAN-based SIMAN (SIMulation ANalysis language) simulation package (Pegden 1982). SIMAN provides a set of FORTRAN-coded commands, called block operations, which are used to build detailed models of continuous or discrete events that make up the operations of any process, such as the operation of an MRS facility. The FACSIM models were designed to run on either an IBM-PC or a VAX minicomputer. The FACSIM/MRS-1 model is flexible enough to collect statistics concerning almost any aspect of the cask receiving and consolidation operations of an MRS facility. The MRS model presently collects statistics on 51 quantities of interest during the simulation. SIMAN reports the statistics with two forms of output: a SIMAN simulation summary and an optional set of SIMAN output files containing data for use by more detailed post processors and report generators

Dry storage of irradiated nuclear fuels and vitrified wastes

International Nuclear Information System (INIS)

Deacon, D.

1982-01-01

A review is given of the work of GEC Energy Systems Ltd. over the years in the dry storage of irradiated fuel. The dry-storage module (designated as Cell 4) for irradiated magnox fuel recently constructed at Wylfa nuclear power station is described. Development work on the long-term dry storage of irradiated oxide fuels is reported. Four different methods of storage are compared. These are the pond, vault, cask and caisson stores. It is concluded that there are important advantages with the passive air-cooled ESL dry stove. (U.K.)

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

International Nuclear Information System (INIS)

Zimmer, A.; Koploy, M.

1991-12-01

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

Safety analysis of casks under extreme impact conditions

International Nuclear Information System (INIS)

Wieser, G.; Qiao, L.; Voelzke, H.; Wolff, D.; Droste, B.

2004-01-01

The determination of the inherent safety of casks under extreme impact conditions has been of increasing interest since the terrorist attacks of 11 September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid-seal system. This can be caused, for example, by a direct aircraft crash (or just its engine) as well as by an impact due to the collapse of a building, e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and-with respect to leak-tightness-relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for finite-element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft or fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like the ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected. (author)

Safety analysis of casks under extreme impact conditions

International Nuclear Information System (INIS)

Wieser, G.; Qiao Linan; Voelzke, H.; Wolff, D.; Droste, B.

2004-01-01

The determination of the inherent safety of casks also under extreme impact conditions has been of increasing interest since the terrorist attacks from 11th September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid seal system. This can be caused e.g. by direct aircraft crash or its engine as well as by an impact due to the collapse of a building e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and - with respect to leak tightness - relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for Finite Element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft and fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected

Spent fuel element storage facility

International Nuclear Information System (INIS)

Ukaji, Hideo; Yamashita, Rikuo.

1981-01-01

Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)

Multi-Purpose Storage Complex description

International Nuclear Information System (INIS)

Nyman, D.H.

1993-01-01

The Multi-Purpose Storage Complex will provide interim storage of radioactive material (irradiated fuel, cesium/strontium capsules, plutonium residuals, canisters of vitrified high-level waste glass, and other radioactive material) at the Hanford Site near Richland, Washington. A Storage Preparation and Shipping Facility is included that will have the capability to stabilize failed metal fuel, segregate high-level solid waste, and package/repackage any of the materials for interim storage/final disposal or subsequent processing. Current technology, both domestic and foreign, will be adapted with the expectation that no new technology will be required. This cost-effective approach will use fuel casks, transport systems, and/or modular vaults that have been licensed in the United States. The complex will have a central control room, and appropriate safeguards and security measures will be incorporated. A specific design objective will be to minimize the amount of secondary waste

Studies and research concerning BNFP: cask handling equipment standardization

International Nuclear Information System (INIS)

McCreery, P.N.

1980-10-01

This report covers the activities of one of the sub-tasks within the Spent LWR Fuel Transportation Receiving, Handling, and Storage program. The sub-task is identified as Cask Handling Equipment Standardization. The objective of the sub-task specifies: investigate and identify opportunities for standardization of cask interface equipment. This study will examine the potential benefits of standardized yokes, decontamination barriers and special tools, and, to the extent feasible, standardized methods and software for handling the variety of casks presently available in the US fleet. The result of the investigations is a compilation of reports that are related by their common goal of reducing cask turnaround time

SCOPE, Shipping Cask Optimization and Parametric Evaluation

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: Given the neutron and gamma-ray shielding requirements as input, SCOPE may be used as a conceptual design tool for the evaluation of various casks designed to carry square fuel assemblies, circular canisters of nuclear waste material, or circular canisters containing 'intact' spent-fuel assemblies. It may be used to evaluate a specific design or to search for the maximum number of full assemblies (or canisters) that might be shipped in a given type of cask. In the 'search' mode, SCOPE will use built-in packing arrangements and the tabulated shielding requirements input by the user to 'design' a cask carrying one fuel assembly (or canister); it will then continue to increment the number of assemblies (or canisters) until one or more of the design limits can no longer be met. In each case (N = 1,2,3...), SCOPE will calculate the steady-state temperature distribution throughout the cask and perform a complete 1-D space/time transient thermal analysis following a postulated half-hour fire; then it will edit the characteristic dimensions of the cask (including fins, if required), the total weight of the loaded case, the steady-state temperature distribution at selected points, and the maximum transient temperature in key components. With SCOPE, the effects of various design changes may be evaluated quickly and inexpensively. 2 - Method of solution: SCOPE assumes that the user has already made an independent determination of the neutron and gamma-ray shielding requirements for the particular type of cask(s) under study. The amount of shielding required obviously depends on the type of spent fuel or nuclear waste material, its burnup and/or exposure, the decay time, and the number of assemblies or canisters in the cask. Source terms (and spectra) for spent PWR and BWR fuel assemblies are provided at each of 17 decay times, along with recommended neutron and gamma-ray shield thicknesses for Pb, Fe, and U-metal casks containing a

Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

Radiation resistant and decontaminable coatings for shipping, interim storage and repository storage casks containing radioactive wastes

International Nuclear Information System (INIS)

Kunze, S.

1995-02-01

All the Corrobesch-DF-Nukelar coatings - black, yellow, blue, red and white - have been excellently decontaminable without and after radiation exposure with 3x10 5 Gy, despite the slightly higher absorbed dose rate applied at KFA Juelich (DIN 55 991 requires ≤1.0 KGy/h). After a further increase to 3x10 6 Gy in the absorbed dose, with an absorbed dose rate up to 1.0 KGy/h conforming to the standard, the coatings black, yellow, blue were still excellent in their decontamination behavior. After exposure to 10 7 Gy all coatings irradiated at Gammaster in their irradiation room (150 m 3 ) with permanent air changes and at absorbed dose rates of 0.9-1.0 KGy/h have been well decontaminable, and the coatings irradiated at KFA Juelich in the 10 l vessel with discontinuous air changes and variable absorbed dose rate (0.22-2.7 KGy/h) have still been fairly well decontaminable only. To be able to evaluate possible changes occurring upon 10 7 Gy radiation exposure, the test specimens were exposed to the action of chemicals according to DIN 55 991 as well as to decontamination cleansing solutions. Different discolorations, very small reductions in brilliancy, and sometimes minor deteriorations in surface hardness occurred. Detrimental visible changes, e.g. bubble and crack formation, swelling, detachment from the base, etc., have not been found for any of the coatings. These results for the test specimens irradiated at Gammaster are identical with the results for the test specimens irradiated at KFA Juelich, except minor deviations. Contrary to expectations, Corrobesch-DF-Nuklear has proved to be a coating material, which, although it consists of organic base material, nevertheless tolerates radiation exposures without visible damage, i.e. conditions under which only electrodeposited nickel coatings have appeared appropriate until now. This means that application of Corrobesch-Nuklear-DF allows the costs of coating of fuel element shipping and storage casks to be reduced

Status of cask procurement strategy to satisfy DOE/OCRWM requirements

International Nuclear Information System (INIS)

Teer, B.R.; Nolan, D.J.; Lake, W.H.

1993-01-01

The Nuclear Waste Policy Act requires the development of a safe and efficient system to transport spent nuclear fuel to and within the Federal Waste Management System. This paper describes the DOE/OCRWM strategy to develop and procure a major component of the Transportation System-the transport cask systems. The original initiative to develop high-capacity innovative designs and its current status is described. The follow-on phase to design and procure proven technology cask systems is also discussed

Shielding properties of protective thin film coatings and blended concrete compositions for high level waste storage packages

International Nuclear Information System (INIS)