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Sample records for resin transport cask

  1. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Abadie, P. [COGEMA Logistics (AREVA Group), Saint-Quentin-en-Yvelines (France)

    2004-07-01

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10{sup 22} at/cm{sup 3} for hydrogen and 9.10{sup 20} at/cm{sup 3} for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C.

  2. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    International Nuclear Information System (INIS)

    Abadie, P.

    2004-01-01

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10 22 at/cm 3 for hydrogen and 9.10 20 at/cm 3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C

  3. Transportation cask contamination weeping

    International Nuclear Information System (INIS)

    Bennett, P.C.; Doughty, D.H.; Chambers, W.B.

    1993-01-01

    This paper describes the problem of cask contamination weeping, and efforts to understand the phenomenon and to eliminate its occurrence during spent nuclear fuel transport. The paper summarizes analyses of field experience and scoping experiments, and concentrates on current modelling and experimental validation efforts. (J.P.N.)

  4. Initiatives in transport cask licensing

    International Nuclear Information System (INIS)

    Patterson, John

    1998-01-01

    The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

  5. 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

  6. 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

  7. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  8. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Kang, Hee Dong; Lee, Heung Young; Seo, Ki Suk; Koo, Jung Hoe; Jung, Sung Hwan; Yoon, Jung Hyun; Lee, Ju Chan; Bang, Kyung Sik; Baek, Chang Yeol

    1992-03-01

    The major goal of this project is to establish the safe transport system and obtain the necessary data for cask development by during research work for the design and safety test of shipping cask. The analysis technique using computer code for design has been studied in the field of structure, thermal and shielding analysis in this study. And also the test and measurement technology was developed for the measuring system of drop and fire test. It is expected that research activity ensured in this job will enable us to ultilize the basic data for the cask development. (Author)

  9. 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

  10. NUHOMS registered - MP197 transport cask

    International Nuclear Information System (INIS)

    Shih, P.; Sicard, D.; Michels, L.

    2004-01-01

    The NUHOMS registered -MP197 cask is an optimized transport design which can be loaded in the spent fuel pool (wet loading) or loaded the canister from the NUHOMS concrete modules at the ISFSI site. With impact limiters attached, the package can be transported within the states or world-wide. The NUHOMS registered -MP197 packaging can be used to transport either BWR or PWR canisters. The NUHOMS registered -MP197 cask is designed to the ASME B and PV Code and meets the requirements of Section III, Division 3 for Transport Packaging. The cask with impact limiters has undergone drop testing to verify the calculated g loadings during the 9m drops. The test showed good correlation with analytical results and demonstrate that the impact limiters stay in place and protect the package and fuel during the hypothetical accidents

  11. 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)

  12. Storage/transport cask design and challenges

    International Nuclear Information System (INIS)

    Houston, J.V.; Viebrock, J.M.

    1989-01-01

    The concept of spent-fuel casks that could be used for both storage and for transport has been around for some years, but was only seriously evaluated when utilities started becoming concerned about adequate fuel storage. In the early 1980s, the U.S. Department of Energy proposed to solve the problem with their away-from-reactor storage facility concept. This was superceded by passage of the Nuclear Waste Policy Act of 1982, which directed the development of one or more waste repositories, the first of which was to be in operation by 1998. Delays in this program now indicate an opening data of 2003 or later. This, together with the lack of significant progress on a monitored retrievable storage facility, leaves the utility companies to solve their storage problems individually. One alternative is to use dual-purpose casks. The use of such a cask should eliminate the need to move the cask and fuel back into the spent-fuel pool for transfer to a transport cask. However, a dual-purpose cask must be licensed for use under both 10CFR71 and 10CFR72 of the U.S. Code of Federal Regulations. The purpose of this paper is to examine the differences between the requirements of 10CFR71 and 10CFR72, to note the changes over the past several years in the NRC's interpretation of 10CFR71 requirements, and to review the design modifications that the Nuclear Assurance Corporation (NAC) believes are required to make a licensed storage cask acceptable for transport under 10CFR71

  13. 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)

  14. Standard casks for the transport of LWR spent fuel

    International Nuclear Information System (INIS)

    Blum, P.

    1986-01-01

    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 have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr

  15. 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

  16. Transport casks help solve spent fuel interim storage problems

    International Nuclear Information System (INIS)

    Dierkes, P.; Janberg, K.; Baatz, H.; Weinhold, G.

    1980-01-01

    Transport casks can be used as storage modules, combining the inherent safety of passive cooling with the absence of secondary radioactive waste and the flexibility to build up storage capacity according to actual requirements. In the Federal Republic of Germany, transport casks are being developed as a solution to its interim storage problems. Criteria for their design and licensing are outlined. Details are given of the casks and the storage facility. Tests are illustrated. (U.K.)

  17. 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.)

  18. Sewage Solids Irradiator Transportation System (SSITS) cask: preliminary design description

    International Nuclear Information System (INIS)

    Eakes, R.G.; Kempka, S.N.; Lamoreaux, G.H.; Sutherland, S.H.

    1983-02-01

    The preliminary design of the Sewage Solids Irradiator Transportation System (SSITS) Cask is presented in this document. The SSITS cask is to be used for the transport of radioactive cesium chloride and strontium fluoride capsules which are of use in irradiators or as heat sources. The SSITS cask is approximately 1.4 m in diameter, 1.3 m high, weighs roughly 9 t, provides 33 cm of steel shielding, and can dissipate up to 5.2 kW of decay heat. The cask design criteria are identified and a description of the cask design and operation is provided. Detailed analyses of the design were performed to demonstrate licensability of the cask by the Nuclear Regulatory Commission (NRC). Results of the analyses indicate that the preliminary design is in compliance with the pertinent regulatory requirements for licensing of a radioactive material transportation container

  19. Cask for concrete shells transportation

    International Nuclear Information System (INIS)

    Labergri, F.

    2001-01-01

    Nowadays, nuclear plant radioactive waste are conditioned in situ into concrete shells. Most of them enter in the industrial waste category defined by the regulations of radioactive material transportation. However, the content of a few ones exceeds the limits set for low specific activity substances. Thus, these shells must be transported into type B packagings. To this end, Robatel has undertaken, for EDF (Electricite de France), the development of a container, named ROBATEL TM R68, for further licensing. The particularity of this packaging is that the lid must have a wide opening to allow the usual handling operations of the concrete shells. This leads to a non-conventional conception, and makes the package more vulnerable to drop test solicitations. In order to define a minimal drop test program on a reduced scale model, we use a simple method to find the most damageable drop angle. (author)

  20. Safety analysis report for packaging: the ORNL loop transport cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Nelms, H.A.; Crowley, W.K.; Just, R.A.

    1977-11-01

    An evaluation of the ORNL loop transport cask demonstrating its compliance with the regulations governing the transportation of radioactive and fissile materials is presented. A previous review of the cask is updated to demonstrate compliance with current regulations, to present current procedures, and to reflect the more recent technology

  1. 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)

  2. 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

  3. 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

  4. 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

  5. Safety analysis report for radwaste foam transport cask

    International Nuclear Information System (INIS)

    Ku, J. H.; Lee, J. C.; Bang, K. S.; Seo, K. S.; Lee, D. W.; Kim, J. H.; Park, S. W.; Lee, J. W.; Kim, K. H.

    1999-08-01

    For the tests and examinations of radwaste foam which generated in domestic nuclear power plants a radioactive material transport cask is needed to transport the radwaste foam from the power plants to KAERI. This cask should be easy to handle in the facilities and safe to maintain the shielding safety of operators. According to the regulations, it should be verified that this cask maintains the thermal and structural integrities under prescribed load conditions by the regulations. The basic structural functions and the integrities of the cask under required load conditions were evaluated. Therefore, it was verified that the cask is suitable to transport radwaste foam from nuclear power plants to KAERI. (author). 11 refs., 10 tabs., 25 figs

  6. 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

  7. Standard casks for the transport of LWR spent fuel

    International Nuclear Information System (INIS)

    Blum, P.

    1985-01-01

    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 have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. 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, - standardisation facilitating fabrication, operation and spare part supply [fr

  8. Developing new transportable storage casks for interim dry storage

    International Nuclear Information System (INIS)

    Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R.

    2004-01-01

    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

  9. 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)

  10. Handbook for structural analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-04-01

    This paper described structural analysis method of radioactive material transport casks for use of a handbook of safety analysis and evaluation. Safety analysis conditions, computer codes for analyses and stress evaluation method are also involved in the handbook. (author)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Issues related to the transport of a transportable storage cask after storage

    International Nuclear Information System (INIS)

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

    1991-01-01

    An evaluation was performed to assess whether the reliability of a transportable storage cask system and the risks associated with its use are comparable to those associated with existing transport cask systems and, if they are not, determine how the transportable storage cask system can be made as reliable as existing systems. Reliability and failure mode analyses of both transport-only casks and transportable storage casks and implementation options are compared. Current knowledge regarding the potential effects of a long-term dry storage environment on spent fuel and cask materials is reviewed. A summary assessment of the consideration for deploying a transportable storage cask (TSC) system with emphasis on preliminary design, validation and operational recommendations for TSC implementations is presented. The analyses conclude that a transportable storage cask can likely be shipped upopened by applying a combination of design considerations and operational constraints, including environmental monitoring and pretransport assessments of functional reliability of the cask. A proper mix of these constraints should yield risk parity with any existing transport cask

  16. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P; Ribeiro, M I; Aparicio, P [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  17. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    Lima, P.; Ribeiro, M.I.; Aparicio, P.

    1998-01-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  18. Contract Report for Usage Inspection of KN-12 Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

  19. Usage Inspection of KN-12 Spent Fuel Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K

    2007-03-15

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse.

  20. Contract Report for Usage Inspection of KN-12 Transport Cask

    International Nuclear Information System (INIS)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K.

    2007-03-01

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse

  1. Usage Inspection of KN-12 Spent Fuel Transport Cask

    International Nuclear Information System (INIS)

    Lee, J. C.; Seo, K. S.; Bang, K. S.; Cho, I. J.; Kim, D. H.; Min, D. K.

    2007-03-01

    The usage inspection of the KN-12 spent nuclear fuel transport package was performed to receive the license for reuse. According to the Korea Atomic Energy Act, all type B transport package should receive and pass the usage inspection every five years. The KN-12 transport cask was designed to transport twelve spent PWR fuel assemblies under wet and dry conditions. The cask was developed and licensed in 2002 in accordance with the Korea and the IAEA's safe transport regulations. The areas of usage inspection include: visual inspection, nondestructive weld inspection, load test, maximum operating pressure test, leakage test, shielding test, thermal test, external surface contamination test. In the results of the usage inspection, the damage or defect could not found out and the performance of the cask was maintained according to the requirements of the regulation. Therefore, the usage inspection was successfully performed to acquire the license for the reuse

  2. STACE: source term analyses for containment evaluations of transport casks

    International Nuclear Information System (INIS)

    Seager, K.D.; Gianoulakis, S.E.; Barrett, P.R.; Rashid, Y.R.; Reardon, P.C.

    1993-01-01

    STACE evaluates the calculated fuel rod response against failure criteria based on the cladding residual ductility and fracture properties as functions of irradiation and thermal environments. The fuel rod gap inventory contains three forms of releasable RAM: (1) gaseous, e.g., 85 Kr, (2) volatiles, e.g., 134 Cs and 137 Cs, and (3) actinides associated with fuel fines. The quantities of these products are limited to that contained within the fuel-cladding gap region and associated interconnected voids. Cladding pinhole failure will also result in the ejection of about 0.003 percent of the fuel, in the form of fines, into the cask cavity. Significant attenuation of the aerosol concentration in the transport cask can occur, depending upon the residence time of the aerosol in the cask compared with its rate of escape from the cask into the environment. (J.P.N.)

  3. 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

  4. High temperature performance limit of containment system of transport cask

    International Nuclear Information System (INIS)

    Kato, Osamu; Saegusa, Toshiari

    1998-01-01

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  5. Underground transportation and handling system for Pollux-casks

    International Nuclear Information System (INIS)

    Schrimpf, C.

    1988-01-01

    The concept for the underground transportation and handling system for Pollux-casks was optimized in a first phase by dividing the process in the repository up into the several transportation and manipulation steps. For each step, the possibilities were described and evaluated by means of a list of criteria (technical, safety and economical criteria). The following concept for the transportation and handling was developed: The casks are transported to the unloading area of the surface facilities by railway or truck. After removal of the transport protection, the entry control is performed. The cask is lifted from the vehicle and placed on a railbound transportation vehicle. This transport unit is transferred to the shaft and placed there ready for shaft hoisting. With the hoisting cage protruding, the transport unit is placed on the hoisting cage by means of a pushing-on device, locked, and then conveyed underground. After arrival on the emplacement level, the transport unit is pulled-off from the hoisting cage and taken over by a mine locomotive and transferred through the transportation and access drifts as far as to the emplacement site. There the locomotive pushed the rail transport vehicle into the emplacement drift, as far as to the designated emplacement position. At the emplacement position, the cask is again lifted by means of hoisting equipment. The rail transport vehicle is pulled out of the emplacement drift and returned to the surface for reloading. After deposition of the cask on the drift floor, the emplacement equipment is pulled back in order to give the operation space free for the slinger backfill truck. Within preceding tests two different backfilling techniques were investigated under realistic conditions: pneumatic backfilling and slinger backfilling. The slinger truck was found to be the most suitable for the designated purpose

  6. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel

    International Nuclear Information System (INIS)

    Bentele, W.; Kinzelmann, T.

    1999-01-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was -2 (limit for surface contamination), presented degrees of contamination >4 Bq cm -2 upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm -2 , as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is negligible. (author)

  7. 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.

  8. 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

  9. 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)

  10. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    International Nuclear Information System (INIS)

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-01-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L R , for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A 2 , are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate

  11. Preliminary design report for the NAC combined transport cask

    International Nuclear Information System (INIS)

    1990-04-01

    Nuclear Assurance Corporation (NAC) is under contract to the United States Department of Energy (DOE) to design, license, develop and test models, and fabricate a prototype cask transportation system for nuclear spent fuel. The design of this combined transport (rail/barge) transportation system has been divided into two phases, a preliminary design phase and a final design phase. This Preliminary Design Package (PDP) describes the NAC Combined Transport Cask (NAC-CTC), the results of work completed during the preliminary design phase and identifies the additional detailed analyses, which will be performed during final design. Preliminary analytical results are presented in the appropriate sections and supplemented by summaries of procedures and assumptions for performing the additional detailed analyses of the final design. 60 refs., 1 fig., 2 tabs

  12. Brittle fracture tests at low temperature for transport cask materials

    International Nuclear Information System (INIS)

    Kosaki, Akio; Ito, Chihiro; Arai, Taku; Saegusa, Toshiari

    1993-01-01

    The IAEA Regulations for the Safe Transport of Radioactive Material were revised in 1985, and brittle fracture assessment at low temperature for transport packages are now required. This report discusses the applicability of the actual method for brittle fracture assessment of type-B transport cask materials used in JAPAN. The necessity of brittle fracture assessment at low temperature was estimated for each material of type-B transport casks used in Japan and the applicability was investigated. Dynamic fracture toughness values, K Id (J Id ), and RT NDT values of Low-Mn Carbon Steels, that are SA 350 Gr.LF1 Modify and SA 516 Gr.70 material which used in type-B transport cask body, were also obtained to check whether or not an easier and conventional test method, that prescribed in ASME CODE SECTION III, can be substituted for the dynamic fracture test method. And for bolt materials, which include 1.8Ni-0.8Cr-0.3Mo Carbon Steel and type 630 H Stainless Steel, toughness data were obtained for reference. (J.P.N.)

  13. Spent fuel transportation cask response to a tunnel fire scenario

    Energy Technology Data Exchange (ETDEWEB)

    Bajwa, C.S. [U.S. Nuclear Regulatory Commission, Washington, D.C. (United States); Adkins, H.E.; Cuta, J.M. [Pacific Northwest National Lab., Richland, WA (United States)

    2004-07-01

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation cask designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the ANSYS {sup registered} and COBRA-SFS computer codes to evaluate the thermal performance of different cask designs. The staff concluded that the transportation casks analyzed would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event. No release of radioactive materials would result from exposure of the casks analyzed to such an event. This paper describes the methods and approach used for this assessment.

  14. Spent fuel transportation cask response to a tunnel fire scenario

    International Nuclear Information System (INIS)

    Bajwa, C.S.; Adkins, H.E.; Cuta, J.M.

    2004-01-01

    On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the effects of this fire on various spent fuel transportation cask designs. The Fire Dynamics Simulator (FDS) code, developed by NIST, was used to determine the thermal environment present in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions in the ANSYS registered and COBRA-SFS computer codes to evaluate the thermal performance of different cask designs. The staff concluded that the transportation casks analyzed would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event. No release of radioactive materials would result from exposure of the casks analyzed to such an event. This paper describes the methods and approach used for this assessment

  15. 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

  16. 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

  17. Neutron and Gamma Shielding Evaluation for KN-12 Spent Nuclear Fuel Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cho, I. J.; Min, D. K.; Lee, J. C.; You, G. S.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, G. H.; Jeong, Y. C.; Ko, Y. W. [Korea Hydro and Nuclear Power Co., LTD., Kori (Korea, Republic of)

    2007-07-01

    The CASTOR KN-12 is designed to transport 12 intact PWR spent fuel assemblies for dry and wet transportation conditions. The overall cask length is 480.1 cm with a wall thickness 37.5 cm. Shield for the KN-12 is maintained by the thick walled cask body and the lid. For neutron shielding, polyethylene rods (PE) are arranged in longitudinal boreholes in the vessel wall and PE-plates are inserted between the cask lid and lid side shock absorber and between the cask bottom and bottom steel plate. The shielding evaluation of the cask has been performed with MCNP to confirm the shielding integrity of cask for pre-service inspection of transport cask.

  18. Design of a transportation cask for irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Nash, K.E.; Gavin, M.E.

    1983-01-01

    A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

  19. Mechanism for transporting used resin

    International Nuclear Information System (INIS)

    Sugimoto, Yoshikazu; Yusa, Hideo; Kamiya, Kunio.

    1975-01-01

    Object: In the operation of a light water reactor type atomic power plant, to permit transport and reuse of used ion exchange resin used for the filtering or cleaning of cooling water or the desalting of radioactive exhaust liquid through an ejector. Structure: Used ion exchange resin within a desalter having high radioactivity is withdrawn through the action of an ejector and led to a solid-liquid separator for separation into used resin and water. The separated resin is directly collected in a storage tank while the separated water is forced through a circulating pump to a gas-liquid separator for separation into gas having radioactivity and water. The separated gas is led to a radioactive gas treatment station while the water deprived of the gas is recirculated by a drive water pump for repeated use. (Kamimura, M.)

  20. Mechanism for transporting used resin

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Y; Yusa, H; Kamiya, K

    1975-01-16

    In the operation of a light water reactor type atomic power plant the objectives is to permit transport and reuse of used ion exchange resin used for the filtering or cleaning of cooling water or the desalting of radioactive exhaust liquid through an ejector. Used ion exchange resin within a desalter having high radioactivity is withdrawn through the action of an ejector and led to a solid-liquid separator for separation into used resin and water. The separated resin is directly collected in a storage tank while the separated water is forced through a circulating pump to a gas-liquid separator for separation into gas having radioactivity and water. The separated gas is led to a radioactive gas treatment station while the water deprived of the gas is recirculated by a drive water pump for repeated use.

  1. Description of from-reactor transportation cask designs

    International Nuclear Information System (INIS)

    Lake, W.H.

    1990-01-01

    This paper describes two from-reactor cask development program contracts. They are a contract for legal weight truck cask designs, and a contract for a rail/barge cask design. The paper also presents several general considerations affecting the cask development program. Two of these which are covered in some detail are the technical topics of burnup credit and source term evaluation

  2. STACE: Source Term Analyses for Containment Evaluations of transport casks

    International Nuclear Information System (INIS)

    Seager, K.D.; Gianoulakis, S.E.; Barrett, P.R.; Rashid, Y.R.; Reardon, P.C.

    1992-01-01

    Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source term has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volatile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking (e.g., the quantity and size distribution of fuel rod breaches) in which experimental validation is planned. The CRUD spallation fraction is the major area where no quantitative data has been found; therefore, this also requires experimental validation. In the interim, STACE conservatively assumes a 100% spallation fraction for computing the releasable activity. The source term methodology also conservatively assumes that there is 1 Ci of residual contamination available for release in the transport cask. However, residual contamination is still by far the smallest contributor to the source term activity

  3. 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

  4. Key technology studies of GY-20 and GY-40 High-capacity cobalt-60 transport casks

    International Nuclear Information System (INIS)

    Liu Huifang; Zhang Xin

    2012-01-01

    GY-20 and GY-40 high-capacity cobalt-60 transport casks are used to transport cobalt-60 industrial irradiators and cobalt-60 bundles. The radioactive contents have special features of high-activity and high residual heat, so only a few countries such as Canada, England and Russia have design capacity. The key technologies and corresponding solutions were studied for the design and manufacture of the cask taking into account the structural, thermal, mechanics and shield requests. A series of tests prove that the cask structure design, design criteria for lead coating structure and quality control measurements are reasonable and effective, and the cask shield integrity can be ensured for all conditions. The casks have ability to transport high-activity sealed sources safely, and the design of cask satisfies the requirement of design code and standard. It can provide reference for other B type package. (authors)

  5. Routine methods for post-transportation accident recovery of spent fuel casks

    International Nuclear Information System (INIS)

    Shappert, L.B.; Pope, R.B.; Best, R.E.; Jones, R.H.

    1991-01-01

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary

  6. CERCA 01: a new safe multi-design MTR transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Faure-Geors, B.S. [Framatome ANP Nuclear Fuel, CERCA, F-26104 Romans (France); Doucet, M.E. [Framatome ANP Nuclear Fuel, F-69006 Lyon (France)

    2001-07-01

    CERCA, a subsidiary company of FRAMATOME ANP, manufactures fuel for research reactors all over the world. To comply with customer requirements, fabrication of material testing reactors elements is a mixed of various parameters. Worldwide transportation of elements requires a flexible cask, which accommodates different designs and meets international transportation regulations. To be able to deliver most of fuel elements, and to cope with non-validation of casks used previously, CERCA decided to design its own cask. All regulatory tests were successfully performed. They completely validated and qualified the safety of this new cask concept. No matter the accidental conditions are, a 5 % {delta}K subcriticality margin is always met.

  7. Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

    2003-01-01

    Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

  8. Transport experience of NH-25 spent fuel shipping cask for post irradiation examination

    International Nuclear Information System (INIS)

    Mori, Ryuji

    1982-01-01

    Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories. Model NH-25 shipping cask was designed, manufactured and oualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2 t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutro n shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell. The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier ''Hinouramaru'' on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated. (author)

  9. 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

  10. Design analysis report for the TN-WHC cask and transportation system

    Energy Technology Data Exchange (ETDEWEB)

    Brisbin, S.A., Fluor Daniel Hanford

    1997-02-13

    This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

  11. 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

  12. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  13. Contamination transfers during fuel transport cask loading. A concrete situation

    International Nuclear Information System (INIS)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G.; Briquet, L.; Baubet, D.

    2002-01-01

    In 1998, a number of contamination cases detected during fuel shipments have been pointed out by the french nuclear safety authority. Wagon and casks external surfaces were partly contaminated upon arrival in Valognes railway terminal. Since then, measures taken by nuclear power plants operators in France and abroad solved the problem. In Germany, a report analyzing the situation in depth has been published in which correctives actions have been listed. In France, EDF launched a large cleanliness program (projet proprete radiologique) in order to better understand contamination transfers mechanisms during power plants exploitation and to list remediation actions to avoid further problems. In this context, CEA Department for Wastes Studies at Cadarache (CEA/DEN/DED) was in charge of a study about contamination transfers during fuel elements loading operations. It was decided to lead experiments for a concrete case. The loading of a transport cask at Tricastin-PWR-1 was followed in november 2000 and different analysis comprising water analysis and smear tests analysis were carried out and are detailed in this paper. Results are discussed and qualitatively compared to those obtained in Philippsburg-BWR, Germany for a similar set of tests. (authors)

  14. 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.

  15. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  16. 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.)

  17. Transport experience with the NAC-1 radioactive materials shipping cask

    International Nuclear Information System (INIS)

    Rollins, J.D.; Hoffman, C.C.

    1976-01-01

    During the first one and one-half years of operation of Nuclear Assurance Corporation's (NAC) four (4) second-generation NAC-1 truck casks, shipments of spent fuel assemblies, fuel rods, and other highly irradiated reactor components have involved over 300,000 cask miles of travel by land, and cask handling at some ten different nuclear facilities. This on-site experience has included the use of various types of auxiliary lifting devices, operational problems with which have identified the need to establish related Quality Assurance procedures in the area of post-fabrication testing. During the course of pre-shipment checkout and testing of the casks minor defects in the upper impact limiter and lower cask shielding wall have been detected and repaired according to procedure. One enroute occurrence with the cask in which an emergency response was implemented has emphasized the need for rigid adherence to procedural checkout before shipment. Periodic inspection and testing are performed as part of the cask license requirement whereby cask components are inspected and/or replaced. During such test periods leaking ball valves and a leaking neutron shield tank have been detected and repaired. (author)

  18. Optimization of cask for transport of radioactive material under impact loading

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Kuldeep, E-mail: kuldeep.brit@gmail.com [Indian Institute of Technology Bombay (India); Pawaskar, D.N.; Guha, Anirban [Indian Institute of Technology Bombay (India); Singh, R.K. [Bhabha Atomic Research Center (India)

    2014-07-01

    Highlights: • Cost and weight are important criteria for fabrication and transportation of cask used for transportation of radioactive material. • Reduction of cask cost by modifying few cask geometry parameters using complex search method. • Maximum von Mises stress generated and deformation after impact as design constraints. • Up to 6.9% reduction in cost and 4.6% reduction in weight observed in the examples used. - Abstract: Casks used for transporting radioactive material need to be certified fit by subjecting them to a specific set of tests (IAEA, 2012). The high cost of these casks gives rise to the need for optimizing them. Conducting actual experiments for the process of design iterations is very costly. This work outlines a procedure for optimizing Type B(U) casks through simulations of the 9 m drop test conducted in ABAQUS{sup ®}. Standard designs and material properties were chosen, thus making the process as realistic as reasonable even at the cost of reducing the options (design variables) available for optimization. The results, repeated for different source cavity sizes, show a scope for 6.9% reduction in cost and 4.6% reduction in weight over currently used casks.

  19. Spent fuel transport cask thermal evaluation under normal and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

    2010-06-15

    The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

  20. Mechanical properties used for the qualification of transport casks

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1993-01-01

    The qualification process that should be sufficient for qualification of a specific cask (material/geometry combination) has been examined. The prototype cask should be tested to determine its overall variation in microstructure, chemistry, and mechanical properties. This prototype may also be subjected to 'proof testing' to demonstrate the validity of the design analysis (including the mechanical properties used in the analysis). The complete mechanical property mapping does not necessarily have to precede the proof testing (i.e., portions of the cask which experience only low (elastic) loads during the drop test are suitable for mechanical test specimens). The behavior of the prototype cask and the production casks are linked by assuring that each cask possesses at least the minimum level of one or more critical mechanical properties. This may be done by measuring the properties of interest directly, or by relying on a secondary measurement (such as subsize mechanical test results or microstructure/compositional measurements) which has been statistically correlated to the critical properties. The database required to show the correlation between the secondary measurement and the valid design property may be established by tests on the material from the prototype cask. The production controls must be demonstrated as being adequate to assure that a uniform product is produced. The testing of coring (or test block or prolongation) samples can only be viewed as providing a valid link to the benchmark results provided by the prototype cask if the process used to create follow-on casks remains essentially similar. The MOSAIK Test Program has demonstrated the qualification method through the benchmarking stage. The program did not establish for qualifying serial production casks through, for example, a correlation between small specimen parameters and valid design fracture toughness properties. Such a correlation would require additional experimental work. (J.P.N.)

  1. 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)

  2. 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

  3. A method for determining the spent-fuel contribution to transport cask containment requirements

    International Nuclear Information System (INIS)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs

  4. 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

  5. Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

    2001-08-15

    Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

  6. 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

  7. First experience in international air transportation of RR SFA in Russian-made TUK-19 casks

    International Nuclear Information System (INIS)

    Kanashov, B.A.; Barinkov, O.P.; Dorofeev, A.N.; Komarov, S.V.; Smirnov, A.V.; Biro, L.; Budu, M.; Ciocanescu, M.

    2010-01-01

    Traditionally, spent fuel assemblies (SFA) have been transported across the Russian Federation by rail in special railcars. New conditions required SFA shipments by other conveyance, i.e. road, sea and even air transport. The air shipment of the VVR-S research reactor SNF in TUK-19 casks from Magurele, Romania in June 2009 was the first experience after new Russian and international regulations for the safe transport of radioactive material came into effect. The preparatory stage of the shipment focused on the issues associated with radiation and nuclear safety both during the loading and transport operations. The project covered development of a technology and equipment for SFA loading into TUK-19 casks and that for the air shipment. The SFAs were loaded into the TUK-19 casks with a specially designed transfer cask, and the SFA-containing packages were transported in specialized freight 20-foot ISO-containers. The safety of the loading and transport operations was ensured both by reliable engineering solutions, and selected conveyances and routes. The paper shows that the loading and the air shipment of the Romanian SFAs in TUK-19 casks does not contradict Romanian, Russian and international regulations for the safe transport of radioactive material. The outcomes of the SNF shipment from Romania confirmed correctness of the solutions and demonstrated high environmental safety. (author)

  8. 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.

  9. Introduction of a new structural material for spent nuclear fuel transportation casks

    International Nuclear Information System (INIS)

    Severson, W.J.; Mello, R.M.; Ciez, A.P.

    1991-01-01

    The From-Reactor Transportation Cask Initiative of the DOE Office of Civilian Radioactive Waste Management (OCRWM) has, since 1988, supported the development of cask systems for the shipment of spent nuclear fuel by both legal weight truck (LWT) and rail or barge. The design basis fuel to be transported would be 10 years out-of-reactor with maximum burnups of 35 and 30 GWD/MTU for PWR and BWR assemblies, respectively. Westinghouse's work on the program led to the development of a common use LWT cask design capable of transporting either three PWR or seven BWR assemblies. This payload in a common use cask is achieved by the use of depleted uranium for the gamma shielding material and Grade 9 titanium as the principal structural material. The use of Grade 9 titanium for cask structures has no certification precedent. This paper describes the work performed to characterize the material and the status of steps taken to gain its acceptance by the NRC, which includes ASME approval of its use in the construction of Section 3 Class 1 components. 9 refs., 7 figs., 9 tabs

  10. 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

  11. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

    2015-01-01

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

  12. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  13. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    International Nuclear Information System (INIS)

    Shah, M.J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 (1). Both the International Atomic Energy Agency regulations for transporting radioactive materials (2, paragraph 727), and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 (3), for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask (3) for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS (4) and LS DYNA (5) finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event

  14. Predicted occurrence rate of severe transportation accidents involving large casks

    International Nuclear Information System (INIS)

    Dennis, A.W.

    1978-01-01

    A summary of the results of an investigation of the severities of highway and railroad accidents as they relate to the shipment of large radioactive materials casks is discussed. The accident environments considered are fire, impact, crash, immersion, and puncture. For each of these environments, the accident severities and their predicted frequencies of occurrence are presented. These accident environments are presented in tabular and graphic form to allow the reader to evaluate the probabilities of occurrence of the accident parameter severities he selects

  15. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  16. 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

  17. Validation of a fracture mechanics approach to nuclear transportation cask design through a drop test program

    International Nuclear Information System (INIS)

    Sorenson, K.B.

    1986-01-01

    Sandia National Laboratories (SNL), under contract to the Department of Energy, is conducting a research program to develop and validate a fracture mechanics approach to cask design. A series of drop tests of a transportation cask is planned for the summer of 1986 as the method for benchmarking and, thereby, validating the fracture mechanics approach. This paper presents the drop test plan and background leading to the development of the test plan including structural analyses, material characterization, and non-destructive evaluation (NDE) techniques necessary for defining the test plan properly

  18. Numerical study of thermal test of a cask of transportation for radioactive material

    International Nuclear Information System (INIS)

    Vieira, Tiago A.S.; Santos, André A.C. dos; Vidal, Guilherme A.M.; Silva Junior, Geraldo E.

    2017-01-01

    In this study numerical simulations of a transport cask for radioactive material were made and the numerical results were compared with experimental results of tests carried out in two different opportunities. A mesh study was also made regarding the previously designed geometry of the same cask, in order to evaluate its impact in relation to the stability of numerical results for this type of problem. The comparison of the numerical and experimental results allowed to evaluate the need to plan and carry out a new test in order to validate the CFD codes used in the numerical simulations

  19. 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)

  20. Study of casks shielded with heavy metal to transport highly radioactive substances

    International Nuclear Information System (INIS)

    Lucchesi, R.F.; Hara, D.H.S.; Martinez, L.G.; Mucsi, C.S.; Rossi, J.L.

    2014-01-01

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

  1. 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)

  2. Radiological Risk Assessment and Cask Materials Qualification for Disposed Sealed Radioactive Sources Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Olteanu, G.; Bujoreanu, D.

    2009-01-01

    The hazardous waste problem imposes to respect national and international agreed regulations regarding their transport, taking into account both for maintaining humans, goods and environment exposure under specified limits, during transport and specific additional operations, and also to reduce impact on the environment. The paper follows to estimate the radiological risk and cask materials qualification according to the design specifications for disposed sealed radioactive sources normal transport situation. The shielding analysis has been performed by using Oak Ridge National Laboratory's SCALE 5 programs package. For thermal analysis and cask materials qualification ANSYS computer code has been used. Results have been obtained under the framework of Advanced system for monitoring of hazardous waste transport on the Romanian territory Research Project which main objective consists in implementation of a complex dual system for on-line monitoring both for transport special vehicle and hazardous waste packages, with data automatic transmission to a national monitoring center

  3. ''MOSAIK {sup registered}'' 20 years of experience with a cask system for transportation, conditioning and storage of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Gestermann, G. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2004-07-01

    In Europe shielding casks for transportation and storage of radioactive waste with higher specific activity are in use. These casks developed by GNS and fabricated from ductile cast iron are called MOSAIK {sup registered} cask. At the moment 4 main types with some subtypes, named MOSAIK {sup registered} I, II, III and 80 T, are in use. The MOSAIK {sup registered} casks are licensed for the German interim storages and the planned final repository MOSAIK {sup registered} II and MOSAIK {sup registered} 80 T are licensed as a type B(U) cask also. Till now more then 4100 MOSAIK {sup registered} cask have been built and loaded with radioactive MAW material.

  4. 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)

  5. 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.))

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  7. A cask fleet operations study

    International Nuclear Information System (INIS)

    1988-03-01

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

  8. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

    2012-01-01

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  9. 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

  10. A new type-B cask design for transporting 252Cf

    International Nuclear Information System (INIS)

    Simmons, C.M.

    2000-01-01

    A project to design, certify, and build a new US Department of Energy (DOE) Type B container for transporting >5 mg of 252 Cf is more than halfway to completion. This project was necessitated by the fact that the existing Oak Ridge National Laboratory (ORNL) Type B containers were designed and built many years ago and thus do not have the records and supporting data that current regulations require. Once the new cask is available, it will replace the existing Type B containers. The cask design is driven by the unique properties of 252 Cf, which is a very intense spontaneous fission neutron source and necessitates a large amount of neutron shielding. The cask is designed to contain up to 60 mg of 252 Cf in the form of californium oxide or californium oxysulfate, in pellet, wire, or sintered material forms that are sealed inside small special-form capsules. The new cask will be capable of all modes of transport (land, sea, and air). The ORNL team, composed of technical and purchasing personnel and using rigorous selection criteria, chose NAC, International (NAC), as the subcontractor for the project. In January 1997, NAC started work on developing the conceptual design and performing the analyses. The original design concept was for a tungsten alloy gamma shield surrounded by two concentric shells of NS-4-FR neutron shield material. A visit to US Nuclear Regulatory Commission (NRC) regulators in November 1997 to present the conceptual design for their comments resulted in a design modification when the question of potential straight-line cracking in the NS-4-FR neutron shield material arose. NAC's modified design includes offset, wedgelike segments of the neutron shield material. The new geometry eliminates concerns about straight-line cracking but increases the weight of the packaging and makes the fabrication more complex. NAC has now completed the cask design and performed the analyses (shielding, structural, thermal, etc.) necessary to certify the cask. The cask

  11. Spent nuclear fuel transportation casks evaluation for water in-leakage

    International Nuclear Information System (INIS)

    Shah, M.J.; Huang, D.T.; Guttmann, J.; Klymyshyn, N.A.; Koeppel, B.J.; Adkins, H.E.

    2004-01-01

    The United States Nuclear Regulatory Commission (USNRC) is responsible for licensing commercial spent fuel storage and transportation systems. To ensure that the regulations are risk-informed, and do not place unnecessary regulatory burden on the industry, the USNRC has been examining its regulations that apply to spent fuel transportation casks for maintaining sub-criticality under hypothetical accident conditions. Code of Federal Regulations, Title 10, Part 71[1] (10 CFR 71), section 71.55(b) requires that, for evaluation of sub-criticality for fissile material packages, water moderation should be assumed to occur to the most reactive credible extent consistent with the chemical and physical form of the contents. This requirement is based on a defense-in-depth policy, and accounts for any possibility of water intrusion into the package. This program is designed to quantify the margins of safety of certified transportation casks to water intrusion following hypothetical accident conditions. This paper describes the current status of analytical work being performed to evaluate two USNRC-certified spent fuel transportation casks, HI-STAR 100[2] and TN-68[3]. The analytical work is performed using the ANSYS registered [4] and LS-DYNA trademark [5] finite element analysis (FEA) codes. The models are sufficiently detailed in the areas of bolt closure interfaces and containment boundaries to evaluate the likelihood water in-leakage under free drop hypothetical accident conditions of 10 CFR 71.73

  12. 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

  13. 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)

  14. 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.

  15. 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)

  16. Transportation cask decontamination and maintenance at the potential Yucca Mountain repository

    International Nuclear Information System (INIS)

    Hartman, D.J.; Miller, D.D.; Hill, R.R.

    1992-04-01

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

  17. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    International Nuclear Information System (INIS)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-01

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

  18. Interactions between cask components and content of packaging for the transport of radioactive material during drop tests

    International Nuclear Information System (INIS)

    Quercetti, T.; Ballheimer, V.; Zeisler, P.; Mueller, K.

    2003-01-01

    This paper describes the analytical, numerical and experimental investigations on the phenomenon of interactions between cask components and content of packages for the transport of radioactive material during drop tests required according to the IAEA Regulations for the Safe Transport of Radioactive Material. Radial and axial gaps between cask components and content are usually necessary for thermal reasons but larger gaps can exist because of the geometrical dimensions of the specified content. Consequently interactions between content and cask components (lid system, cask body, etc.) are possible and can not be excluded during drop tests. Interactions in this context are relative movements between cask and content which are mainly due to elastic spring effects after releasing the cask for the free drop. These relative movements can cause interior collisions between content and cask during the main impact of the package onto the unyielding target. Drop tests with various types of Type A and Type B packages fully instrumented with strain gauges and accelerometers showed that these interactions respectively interior collisions can be considerable relating to high forces acting on cask lids, lid bolts and the content. Of course the real quantitative consequences of the interactions depend upon different conditions, among others the drop orientation, the design characteristics of the impact limiters, the dimensions of the gaps, the material characteristics of the contents, etc. . In order to investigate more precisely the phenomenon of interactions BAM carried out finite element calculations for the named casks using the ABAQUS/ Standard and ABAQUS/ Explicit computer code comparing them with results obtained from experiments. Additionally, tests with a simplified model instrumented with accelerometers were carried out accompanied by finite element calculations and analytical calculations using MATHEMATICA. The investigations on the mentioned phenomena of interaction

  19. Performance of CASTOR {sup registered} HAW cask cold trials for loading, Transport and storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Horn, Thomas; Graf, Wilhelm [GNS Gesellschaft fuer Nuklear-Service mbH (Germany)

    2009-07-01

    With over 30 years of experience in the design, manufacturing, assembly and loading of CASTOR {sup registered} casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high active wastes (meanwhile over 1.000 casks loaded and stored and more than 1.500 ordered). GNS's products are used at around 30 sites worldwide for a wide range of inventories from pressurised and boiling water reactor fuels (PWR, VVER and BWR, RBMK), thorium high-temperature reactor fuels (THTR) and research reactor fuels (MTR) to vitrified high active wastes (HAW) from reprocessing plants. GNS is responsible for all nuclear wastes resulting from German Nuclear Power Plants and assists and/or performs in the loading and dispatch of CASTOR {sup registered} casks as well as their transport to and storage at central interim storage facilities and local interim storage areas. (orig.)

  20. 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

  1. 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

  2. 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

  3. 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

  4. FINCRUSH : a computer program for impact analysis of radioactive material transport cask with fins

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1997-05-01

    In drop impact analyses for radioactive material transport cask with cooling fins, relationship between fin plastic deformation and fin energy absorption is used. This relationship was obtained by ORNL experiments and MONSER Co. in Canada. Based on ORNL experiments, a computer program FINCRUSH has been developed for rapid safety analysis of cask drop impact to obtain the maximum impact acceleration and the maximum fin deformation. Main features of FINCRUSH are as follows: (1) annulus fins on a cylindrical shell and plate fins on a disk can be treated, (2) it is capable of graphical representations for calculation results and fin absorption energy data and (3) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of the FINCRUSH. In the paper, brief illustration of calculation method of FINCRUSH is presented. The second section presents comparisons between FINCRUSH and experimental results. The third section provides a use's guide for FINCRUSH. (author)

  5. FINCRUSH : a computer program for impact analysis of radioactive material transport cask with fins

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-05-01

    In drop impact analyses for radioactive material transport cask with cooling fins, relationship between fin plastic deformation and fin energy absorption is used. This relationship was obtained by ORNL experiments and MONSER Co. in Canada. Based on ORNL experiments, a computer program FINCRUSH has been developed for rapid safety analysis of cask drop impact to obtain the maximum impact acceleration and the maximum fin deformation. Main features of FINCRUSH are as follows: (1) annulus fins on a cylindrical shell and plate fins on a disk can be treated, (2) it is capable of graphical representations for calculation results and fin absorption energy data and (3) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of the FINCRUSH. In the paper, brief illustration of calculation method of FINCRUSH is presented. The second section presents comparisons between FINCRUSH and experimental results. The third section provides a use`s guide for FINCRUSH. (author)

  6. 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

  7. 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

  8. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

    1992-01-01

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both nominal use and accident conditions to serve the dual-role of shielding and containment, the use of other structure materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describesa two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature. The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracmm resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as win be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term ''structural credit'' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.)

  9. Incentives for the use of depleted uranium alloys as transport cask containment structure

    International Nuclear Information System (INIS)

    McConnell, P.; Salzbrenner, R.; Wellman, G.W.; Sorenson, K.B.

    1993-01-01

    Radioactive material transport casks use either lead or depleted uranium (DU) as gamma-ray shielding material. Stainless steel is conventionally used for structural containment. If a DU alloy had sufficient properties to guarantee resistance to failure during both normal use and accident conditions to serve the dual-role of shielding and containment, the use of other structural materials (i.e., stainless steel) could be reduced. (It is recognized that lead can play no structural role.) Significant reductions in cask weight and dimensions could then be achieved perhaps allowing an increase in payload. The mechanical response of depleted uranium has previously not been included in calculations intended to show that DU-shielded transport casks will maintain their containment function during all conditions. This paper describes a two-part study of depleted uranium alloys: First, the mechanical behavior of DU alloys was determined in order to extend the limited set of mechanical properties reported in the literature (Eckelmeyer, 1991). The mechanical properties measured include the tensile behavior the impact energy. Fracture toughness testing was also performed to determine the sensitivity of DU alloys to brittle fracture. Fracture toughness is the inherent material property which quantifies the fracture resistance of a material. Tensile strength and ductility are significant in terms of other failure modes, however, as will be discussed. These mechanical properties were then input into finite element calculations of cask response to loading conditions to quantify the potential for claiming structural credit for DU. (The term 'structural credit' describes whether a material has adequate properties to allow it to assume a positive role in withstanding structural loadings.) (J.P.N.)

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

    Directory of Open Access Journals (Sweden)

    Rezaeian Mahdi

    2015-01-01

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

  11. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    Lee, Dew Hey; Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong

    2002-03-01

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  12. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    Lee, Dew Hey; Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Ho Chul; Hong, Song Jin; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun

    2001-03-01

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS is completely corresponded. And The integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  13. Operations of the LR56 radioactive liquid cask transport system at U.S. Department of Energy sites

    International Nuclear Information System (INIS)

    Davidson, J.S.; Hornstra, D.J.; Sazawal, V.K.; Clement, G.

    1996-01-01

    The LR56 cask system is licensed for use in France under Certificate of Compliance F/309/B(U)F for transport of 4,000-liter volumes of radioactive liquids. Three LR56 cask systems (with modifications for use at Department of Energy (DOE) sites) have been purchased for delivery at the Hanford Site, Oak Ridge National Laboratory (ORNL), and Savannah River Site (SRS). The LR56 cask systems will be used for on-site transfers of Type B quantities of radioactive liquid waste. The ORNL unit will also be used as a Type A packaging for transfers of radioactive liquids between DOE sites. This paper discusses LR56 operating features and the use of the cask system at the three DOE sites

  14. Safety assessment technology on the free drop impact and puncture analysis of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Ho Chul; Hong, Song Jin; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-03-15

    In this study, the regulatory condition and analysis condition is analyzed for the free drop and puncture impact analysis to develop the safety assessment technology. Impact analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. LS-DYNA3D and ABAQUS is suitable for the free drop and the puncture impact analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS is completely corresponded. And The integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  15. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear and Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong [Chungnam National Univ., Taejon (Korea, Republic of)

    2002-03-15

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  16. Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC

    International Nuclear Information System (INIS)

    West, L.; Akin, B.P.; Lemley, E.C.

    1995-01-01

    Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations

  17. A program to qualify ductile cast iron for use as a containment material for type B transport cask

    International Nuclear Information System (INIS)

    Golliher, K.G.; Sorenson, K.B.; Witt, C.R.

    1990-01-01

    This paper reports on the Department of Energy (DOE) investigations for the use of ductile cast iron (DCI) as a candidate material for radioactive material transportation cask construction. The investigation will include materials testing and full-scale cask testing. The major effort will focus on materials qualification and cask evaluation of the 9 meter and puncture drop test events. Interaction by contract with the private industry, the American Society for Testing and Materials (ASTM) Committee A4.04, and the Electric Power Research Institute (EPRI) will be actively pursued to establish material specification acceptance criteria for ductile iron use as a cask material in the United States of America (USA). All test results will be documented in the safety analysis report for packaging for submission to the U.S. Nuclear Regulatory Commission (NRC). The goal of this program is a certificate of compliance for DCI from the NRC to transport high-level radioactive materials. The acceptance of DCI within the USA cask design community will offer an alternative to present-day materials for cask construction, and its entry has the potential of providing significant cost-savings

  18. Below Grade Assessment of Spent Nuclear Fuel Cask Transport Route

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    1999-01-01

    The report provides an assessment of the route for the SNF Fuel transport system from the K Basins to the CVDF and to the CSB. Results include the identification of any underground structures or utilities traveled over by the transport, the overburden depths for all locations identified, evaluation of the loading conditions, and determination of the effects of the loads on the structures and utilities

  19. 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

  20. 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)

  1. Impacts of transportation regulations on spent fuel and high level waste cask design

    International Nuclear Information System (INIS)

    Lake, W.H.

    1992-01-01

    The regulation of radioactive material transport has a long and successful history. Prior to 1966, these activities were regulated by the Interstate Commerce Commission (ICC) Bureau of Explosives (the ICC was predecessor to the Department of Transportation (DOT)). In 1966, the Atomic Energy Commission (AEC) developed what is now 10 CFR 71, concurrently with the development of similar international standards. In 1975, the AEC was reorganized and the Nuclear Regulatory Commission (NRC) was established as an independent regulatory commission. The NRC was given responsibility for the regulation of commercial use of radioactive materials, including transportation. This paper discusses various aspects of the NRC's role in the transport of radioactive material as well as its role in the design and certification of casks necessary to the transport of this material

  2. 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

  3. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ''fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions

  4. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  5. Development of casting technology and localization for a medical radioisotope transport cask

    International Nuclear Information System (INIS)

    Lee, Y. S.; Kim, H. S.; Jang, S. J.; Seo, K. S.; Kim, C. G.

    2003-01-01

    In order to localize the shielding casks for shipping medical isotopes, this research was carried out. The various casting factors such as the diameter of shielding casting, the temperature of melt and the temperature gradient of a mold were determined with the calculation results of solidification analysis computer code. Through the experiment, the manufacturing method of Ti core was developed to have no defects causing casting failure. As a results of casting experiment, depleted uranium shielding castings were successfully cast without any defect. Also as the results of the radiation shielding capability test, it was good enough to satisfied the standards of transport regulations

  6. Soviet-German cooperation in safety improvements of spent fuel transport casks

    International Nuclear Information System (INIS)

    Schulz-Forberg, B.; Zeisler, P.; Droste, B.; Kondratiev, A.; Kozlov, Ju.; Tichinov, N.

    1993-01-01

    The paper gives a survey of the Soviet (Russian)-German activities which started in 1988 with the objective of creating a long-term scientific-technical cooperation in the field of transport and storage casks for spent nuclear fuel. The first step, i.e., the step of informing each other about the state of development is done. The more complicated second phase with concerted common activities of both the Russian and German competent Authorities and industrial enterprises is intended to start the in near future. (author)

  7. Human factors engineering applications in the testing of the legal weight truck cask transportation system

    International Nuclear Information System (INIS)

    Smith, T.C.; Peck, M. III; Sealock, R.A.

    1994-01-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) will collect performance data to be used in limited human factors engineering analysis of the light weight tractor as a component of the legal weight truck cask transport system. The Management and Operating contractor will provide an analysis and comparison of limited data on driver behavior and subjective driver evaluations of the light weight tractor performance versus that of a heavier baseline tractor. A significant difference in performance data would suggest that given tractor configurations affect driver behavior differently

  8. 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)

  9. 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

  10. Review of the Effects of Normal Conditions of Transport on Spent Fuel Integrity in Transportation Casks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Junggoo; Yoo, Youngik; Lee, Seongki; Lim, Chaejoon [Korea Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2014-10-15

    Spent fuel(SF) storage capacity of each domestic nuclear power plant will reach a saturated state in the near future. Although there are several methods of SF disposal, interim storage is suggested as the most realistic and promising alternative. SF integrity evaluation is a regulatory requirement that is described in Part 71 of Code of Federal Regulations, Title 10 of the U..S. NRC licensing requirement. In this paper, the report is reviewed written by EPRI in US and it is helpful to a development of domestic SF integrity evaluation technology. EPRI report about integrity evaluation method on normal conditions of high burn-up spent fuel transport is reviewed. First, dynamic forces occurred in one-foot side drop are calculated. And deformation patterns and fuel rods responses by dynamic forces calculated from spent fuel and cask model are analyzed. It is shown that the damage of fuel rods is not occurred by the dynamic forces on normal conditions. Assembly distortion is not predicted, by virtue of the facts that the spacer grids do not experience significant permanent deformation. Axial forces, bending moments and pinch forces of fuel rods are calculated and compared with the results under the hypothetical accident conditions. No occurrence of transverse tearing mode that is the most serious damage mode in side drop case is predicted. Till now, in Korea, regulatory requirements related with structural integrity of spent fuel are not specified such as 10CFR71. To establish own regulation standards, producing and analyzing sufficient experimental data must be performed preferentially. Based on this, failure analysis and criteria establishment are necessary through modeling and analyzing of spent fuel.

  11. 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

  12. 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

  13. Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

    2000-07-01

    Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

  14. A risk-informed basis for establishing non-fixed surface contamination limits for spent fuel transportation casks

    International Nuclear Information System (INIS)

    Rawl, R.R.; Eckerman, K.F.; Bogard, J.S.; Cook, J.R.

    2004-01-01

    The current limits for non-fixed contamination on packages used to transport radioactive materials were introduced in the 1961 edition of the International Atomic Energy Agency (IAEA) transport regulations and were based on radiation protection guidance and practices in use at that time. The limits were based on exposure scenarios leading to intakes of radionuclides by inhalation and external irradiation of the hands. These considerations are collectively referred to as the Fairbairn model. Although formulated over 40 years ago, the model remains unchanged and is still the basis of current regulatory-derived limits on package non-fixed surface contamination. There can also be doses that while not resulting directly from the contamination, are strongly influenced by and attributable to transport regulatory requirements for contamination control. For example, actions necessary to comply with the current derived limits for light-water-reactor (LWR) spent nuclear fuel (SNF) casks can result in significant external doses to workers. This is due to the relatively high radiation levels around the loaded casks, where workers must function during the measurement of contamination levels and while decontaminating the cask. In order to optimize the total dose received due to compliance with cask contamination levels, it is necessary to take into account all the doses that vary as a result of the regulatory limit. Limits for non-fixed surface contamination on spent fuel casks should be established by using a model that considers and optimizes the appropriate exposure scenarios both in the workplace and in the public environment. A risk-informed approach is needed to ensure optimal use of personnel and material resources for SNF-based packaging operations. This paper is a summary of a study sponsored by the US Nuclear Regulatory Commission and performed by Oak Ridge National Laboratory that examined the dose implications for removable surface contamination limits on spent fuel

  15. 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

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

    International Nuclear Information System (INIS)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

    2004-01-01

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

  17. 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

  18. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1997-12-01

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson's ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user's guide for computer program and input data for the IMPACLIB. (author)

  19. IMPACLIB: a material property data library for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-12-01

    The paper describes the structural data library and graphical program for impact and stress analyses of radioactive material transport casks. Four kinds of material data, structure steels, stainless steels, leads and woods are compiled. These materials are main structural elements of casks. Structural data such as, coefficient of thermal expansion, modulus of longitudinal elasticity, modulus of transverse elasticity, Poisson`s ratio and stress-strain relationship have been tabulated. Main features of IMPACLIB are as follows: (1) data have been tabulated against temperature or strain rate, (2) thirteen kinds of polynominal fitting for stress-strain curve are available, (3) it is capable of graphical representations for structural data and (4) the IMPACLIB is able to be used on not only main frame computers but also work stations (OS UNIX) and personal computers (OS Windows 3.1). In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides a user`s guide for computer program and input data for the IMPACLIB. (author)

  20. CRUSH1: a simplified computer program for impact analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-07-01

    In drop impact analyses for radioactive transport casks, it has become possible to perform them in detail by using interaction evaluation, computer programs, such as DYNA2D, DYNA3D, PISCES and HONDO. However, the considerable cost and computer time are necessitated to perform analyses by these programs. To meet the above requirements, a simplified computer program CRUSH1 has been developed. The CRUSH1 is a static calculation computer program capable of evaluating the maximum acceleration of cask bodies and the maximum deformation of shock absorbers using an Uniaxial Displacement Method (UDM). The CRUSH1 is a revised version of the CRUSH. Main revisions of the computer program are as follows; (1) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows 3.1 or Windows NT) are available for use of the CRUSH1 and (2) input data set are revised. In the paper, brief illustration of calculation method using UDM is presented. The second section presents comparisons between UDM and the detailed method. The third section provides a use`s guide for CRUSH1. (author)

  1. CRUSH1: a simplified computer program for impact analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1996-07-01

    In drop impact analyses for radioactive transport casks, it has become possible to perform them in detail by using interaction evaluation, computer programs, such as DYNA2D, DYNA3D, PISCES and HONDO. However, the considerable cost and computer time are necessitated to perform analyses by these programs. To meet the above requirements, a simplified computer program CRUSH1 has been developed. The CRUSH1 is a static calculation computer program capable of evaluating the maximum acceleration of cask bodies and the maximum deformation of shock absorbers using an Uniaxial Displacement Method (UDM). The CRUSH1 is a revised version of the CRUSH. Main revisions of the computer program are as follows; (1) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows 3.1 or Windows NT) are available for use of the CRUSH1 and (2) input data set are revised. In the paper, brief illustration of calculation method using UDM is presented. The second section presents comparisons between UDM and the detailed method. The third section provides a use's guide for CRUSH1. (author)

  2. 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)

  3. Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

    International Nuclear Information System (INIS)

    Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

    2000-01-01

    According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

  4. 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

  5. Estimation of gamma dose rate from hulls and shield design for the hull transport cask of Fuel Reprocessing Plant (FRP)

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    In Fuel Reprocessing Plant (FRP), un-dissolved clad of fuel pins known as hulls are the major sources of high level solid waste. Safe handling, transport and disposal require the estimation of radioactivity as a consequent of gamma dose rate from hulls in fast reactor fuel reprocessing plant in comparison with thermal reactor fuel. Due to long irradiation time and low cooling of spent fuel, the evolution of activation products 51 Cr, 58 Co, 54 Mn and 59 Fe present as impurities in the fuel clad are the major sources of gamma radiation. Gamma dose rate from hull container with hulls from Fuel Sub Assembly (FSA) and Radial Sub Assembly (RSA) of Fuel Reprocessing Plant (FRP) was estimated in order to design the hull transport cask. Shielding computations were done using point kernel code, IGSHIELD. This paper describes the details of source terms, estimation of dose rate and shielding design of hull transport cask in detail. (author)

  6. The buckling of fuel rods in transportation casks under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Bjorkman, G.S.

    2004-01-01

    The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

  7. Opportunities in the certification of metal storage casks for spent fuel transport

    International Nuclear Information System (INIS)

    Teer, B.R.

    1988-01-01

    The current regulatory climate in the Transportation Certification Branch of the Nuclear Regulatory Commission is such that the applicant for a Certificate of Compliance must submit a well prepared, comprehensive Safety Analysis Report which meets all requirements of Regulatory Guide. All methods and procedures for structural, nuclear and thermal analyses must be qualified and benchmarked. The materials used for the containment boundary, impact limiters, neutron and gamma shields and internal structures must have documented and reproducible properties and, ideally, be referred in a recognized national standard. And, finally, in most cases, the analyses must be supplemented by model or full scale test results. The paper discusses some recent Transnuclear experiences with the NRC Transportation Certification Branch in attempting to get the TN-BRP and TN-REG casks certified for the transport of spent fuel. The authors have utilized non-standard materials for the baskets and a modified material for the containment. The impact limiters are not of a conventional design. Some of the analytical procedures are proprietary codes which are not used by others in the industry

  8. A robotic system to conduct radiation and contamination surveys on nuclear waste transport casks

    International Nuclear Information System (INIS)

    Harrigan, R.W.; Sanders, T.L.

    1990-06-01

    The feasibility of performing, numerous spent fuel cask operations using fully integrated robotic systems is under evaluation. Using existing technology, operational and descriptive software and hardware in the form of robotic end effectors are being designed in conjunction with interfacing cask components. A robotic radiation and contamination survey system has been developed and used on mock-up cask hardware to evaluate the impact of such fully automated operations on cask design features and productivity. Based on experience gained from the survey system, numerous health physics operations can be reliably performed with little human intervention using a fully automated system. Such operations can also significantly reduce time requirements for cask-receiving operations. 7 refs., 51 figs., 6 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  10. The application of fracture mechanics to the safety assessment of transport casks for radioactive material

    International Nuclear Information System (INIS)

    Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

    2004-01-01

    BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety 'proof of package' for components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new 'Guidelines for the application of ductile cast iron for transport and storage casks for radioactive materials'. Based on these guidelines, higher stresses than before will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's advisory material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will conclude the paper. (author)

  11. Peak cladding temperature in a spent fuel storage or transportation cask

    International Nuclear Information System (INIS)

    Li, J.; Murakami, H.; Liu, Y.; Gomez, P.E.A.; Gudipati, M.; Greiner, M.

    2007-01-01

    From reactor discharge to eventual disposition, spent nuclear fuel assemblies from a commercial light water reactor are typically exposed to a variety of environments under which the peak cladding temperature (PCT) is an important parameter that can affect the characteristics and behavior of the cladding and, thus, the functions of the spent fuel during storage, transportation, and disposal. Three models have been identified to calculate the peak cladding temperature of spent fuel assemblies in a storage or transportation cask: a coupled effective thermal conductivity and edge conductance model developed by Manteufel and Todreas, an effective thermal conductivity model developed by Bahney and Lotz, and a computational fluid dynamics model. These models were used to estimate the PCT for spent fuel assemblies for light water reactors under helium, nitrogen, and vacuum environments with varying decay heat loads and temperature boundary conditions. The results show that the vacuum environment is more challening than the other gas environments in that the PCT limit is exceeded at a lower boundary temperature for a given decay heat load of the spent fuel assembly. This paper will highlight the PCT calculations, including a comparison of the PCTs obtained by different models.

  12. Mechanical properties used for the qualification of transport casks: Prototype development and extension to serial production

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1991-01-01

    A thorough understanding of the mechanical behavior of material in a specific cask is required to properly analyze the structural response of the cask. An appropriate way to establish this understanding is through laboratory testing of cask material. The laboratory testing that was done to support the MOSAIK Drop Test Program is summarized as an example of how mechanical properties can be mapped for a prototype cask. The broad range behavior to be understood. This is necessary for the proper application of fracture mechanics, and focuses on fracture toughness as the inherent materials property which quantifies the fracture resistance of a material. The understanding established by a mechanics to a particular prototype, behavior of a prototype must be correctly associated with parameters which can be measured on production casks. Since the production casks cannot be destructively tested, measurements are commonly made on sub-size specimens. This may prevent direct measurement of valid design properties. An additional database may then be required to establish the correlation between sub-size specimen measurements and valid design properties. This is illustrated by outlining the additional testing which would be necessary to allow the successful verification of the MOSAIK Drop Test Program to be extended from the prototype to serially produced casks

  13. Mechanical properties used for the qualification of transport casks: Prototype development and extension to serial production

    International Nuclear Information System (INIS)

    Salzbrenner, R.; Crenshaw, T.B.; Sorenson, K.B.

    1992-01-01

    A thorough understanding of the mechanical behavior of material in a specific cask is required to properly analyze the structural response of the cask. An appropriate way to establish this understanding is through laboratory testing of cask material. The laboratory testing that was done to support the MOSAIK Drop Test Program is summarized as an example of how mechanical properties can be mapped for a prototype cask. The broad range of measured properties allows the critical aspects of mechanical behavior to be understood. This is necessary for the proper application of fracture mechanics, and focuses on fracture toughness as the inherent materials property which quantifies the fracture resistance of a material. The general fracture mechanics approach and its application to specific cask designs are described elsewhere (Salzbrenner et al. 1990, Sorenson et al. 1992a, Sorenson et al. 1992b). The understanding established by a thorough mapping of the mechanical properties is necessary to apply fracture mechanics to a particular prototype, but it is not sufficient for qualifying serially produced casks. The mechanical behavior of a prototype must be correctly associated with parameters which can be measured on production casks. Since the production casks cannot be destructively tested, measurements are commonly made on sub-size specimens. This may prevent direct measurement of valid design properties. An additional database may then be required to establish the correlation between sub-size specimen measurements and valid design properties. This is illustrated by outlining the additional testing which would be necessary to allow the successful verification of the MOSAIK Drop Test Program to be extended from the prototype to serially produced casks

  14. Multi-wall cask advantages with quarter-scale model drop test results for the NAC-STC Storable Transport Cask

    International Nuclear Information System (INIS)

    Danner, T.A.; Thompson, T.C.; Yaksh, M.C.

    1993-01-01

    Physical drop test results for a quarter-scale model multi-wall cask are presented for the 9 meter end, side and oblique drops with impact limiters, and for the 1 meter side and closure lid pin puncture drops. Lessons learned and final cask test qualification are presented. (J.P.N.)

  15. 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

  16. Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Schneider, J. P.

    2010-01-01

    Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

  17. Compliant tiedown system for securing a cask for transporting radioactive materials to a vehicle

    International Nuclear Information System (INIS)

    Richardson, R.V.

    1991-01-01

    This patent describes a tiedown system for releasably securing a cask having opposing trunnions to a cradle assembly that in turn is mounted onto the deck and frame of a vehicle. The cradle assembly includes at least one cradle member for supporting the weight of the cask, and wherein the cask trunnions are displaceable in opposite vertical directions relative to the cradle member of the cradle assembly when the assembly is torsionally flexed, and the cask rotates relative to the cradle member, about a horizontally oriented axis comprising a compliant linkage means including a tie-bar means operably connected between the trunnions and moveably mounted within the cradle assembly for applying a biasing force on each of the trunnions that remains substantially constant despite the opposite vertical displacement of the trunnions relative to the cradle member that results from the torsional flexing of the vehicle deck

  18. Force reconstruction for the slapdown test of a nuclear transportation cask

    International Nuclear Information System (INIS)

    Bateman, V.I.; Carne, T.G.; Gregory, D.L.; Attaway, S.W.; Yoshimura, H.R.

    1989-01-01

    Two force reconstruction techniques were used to evaluate the slapdown response of a 1/3 scale model solid steel, spent fuel cask dropped 30 ft onto an unyielding target. The two techniques are: the sum of weighted acceleration technique (SWAT) and the deconvolution technique (DECON). A brief description and the calibration of the techniques as applied to the cask are presented. For the slapdown test, both techniques yielded very similar resultant forces and provided more accurate definition of the force-time history for the cask than is available from conventional data reduction methods. An applied moment, a measurement previously unobtainable from conventional cask accelerometer data reduction techniques, was determined with SWAT. The angular velocity calculated with SWAT was verified with photometric measurements. 9 refs., 22 figs

  19. CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-10-01

    A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  20. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1998-03-01

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user's guide for computer program and input data for THERMLIB. (author)

  1. THERMLIB: a material property data library for thermal analysis of radioactive material transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The paper describes an heat conduction data library and graphical program for analysis of radioactive material transport casks. More than 1000 of material data are compiled in the data library which was produced by Lawrence Livermore Laboratory. Thermal data such as, density, thermal conductivity, specific heat, phase-change or solid-state, transition temperature and latent heat have been tabulated. Using this data library, a data library processing program THERMLIB for thermal analysis has been developed. Main features of THERMLIB are as follows: (1) data have been tabulated against temperature, (2) more than 1000 material data are available, (3) it is capable of graphical representations for thermal data and (4) not only main frame computer but also work stations (OS UNIX) and personal computer (OS Windows) are available for use of THERMLIB. In the paper, brief illustration of data library is presented in the first section. The second section presents descriptions of structural data. The third section provides an user`s guide for computer program and input data for THERMLIB. (author)

  2. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-01-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  3. On the requirement for remodelling the spent nuclear fuel transportation casks for research reactors. A review of the drop impact analyses of JRC-80Y-20T

    International Nuclear Information System (INIS)

    2005-07-01

    The Japan Atomic Energy Research Institute (JAERI) constructed two stainless steel transportation casks, JRC-80Y-20T, for spent nuclear fuels of research reactors and had utilized them for transportation since 1981. A modification of the design was applied to the United States of America (USA) for transportation of silicide fuels. Additional analyses employing the impact analysis code LS-DYNA that was often used for safety analysis were submitted by the JAERI to the USA in 2003 to show integrity of the packages; the casks were still not approved, because inelastic deformation was occurred on the surface of the lid touching to the body. To resolve this problem on design approval of transportation casks, a review group was formed in June 2004. The group examined the impact analyses by reviewing the input data and performing the sensitivity analyses. As the drop impact analyses were found to be practically reasonable, it was concluded that the approval of the USA for the transportation casks could not be obtained just by revising the analyses; therefore, remodelling the casks is required. (author)

  4. 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

  5. Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-10-01

    The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

  6. Latching device for securing a closure to a cask for transporting radioactive waste

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Cruz, R.R.; Bieberbach, G.

    1991-01-01

    This patent describes a latching device for removably securing and sealingly engaging a closure around an opening in a cask defined by an edge of the cask. It comprises: at least three shear key assemblies uniformly spaced around the outer portion of the closure, each of which includes a shear key means having a bolt portion movably mounted in the closure, and a latch portion that is insertable into and retractable out of a slot means in the cask edge in response to the application of a closing and an opening force applied to the bolt portion, at least one drive mechanism located within a cavity in the closure and connected to the bolt portion of the shear key assemblies for applying the closing and opening forces to the bolt portions

  7. Development of neutron shielding material for cask

    International Nuclear Information System (INIS)

    Najima, K.; Ohta, H.; Ishihara, N.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    Since 1980's Mitsubishi Heavy Industries, Ltd (MHI) has established transport and storage cask design 'MSF series' which makes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed neutron shielding material. This neutron shielding material has been developed for improving durability under high condition for long term. Since epoxy resin contains a lot of hydrogen and is comparatively resistant to heat, many casks employ epoxy base neutron shielding material. However, if the epoxy base neutron shielding material is used under high temperature condition for a long time, the material deteriorates and the moisture contained in it is released. The loss of moisture is in the range of several percents under more than 150 C. For this reason, our purpose was to develop a high durability epoxy base neutron shielding material which has the same self-fire-extinction property, high hydrogen content and so on as conventional. According to the long-time heating test, the weight loss of this new neutron shielding material after 5000 hours heating has been lower than 0.04% at 150 C and 0.35% at 170 C. A thermal test was also performed: a specimen of neutron shielding material covered with stainless steel was inserted in a furnace under condition of 800 C temperature for 30 minutes then was left to cool down in ambient conditions. The external view of the test piece shows that only a thin layer was carbonized

  8. Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

    International Nuclear Information System (INIS)

    Lopes, Pedro; Vale, Alberto; Ventura, Rodrigo

    2015-01-01

    The Cask and Plug Remote Handling System (CPRHS) and the respective Cask Transfer System (CTS) are designed to transport activated components between the reactor and the hot cell buildings of ITER during maintenance operations. In nominal operation, the CPRHS/CTS shall operate autonomously under human supervision. However, in some unexpected situations, the automatic mode must be overridden and the vehicle must be remotely guided by a human operator due to the harsh conditions of the environment. The CPRHS/CTS is a rhombic-like vehicle with two independent steerable and drivable wheels along its longitudinal axis, giving it omni-directional capabilities. During manual guidance, the human operator has to deal with four degrees of freedom, namely the orientations and speeds of two wheels. This work proposes a Human Machine Interface (HMI) to manage the degrees of freedom and to remotely guide the CPRHS/CTS in ITER taking the most advantages of rhombic like capabilities. Previous work was done to drive each wheel independently, i.e., control the orientation and speed of each wheel independently. The results have shown that the proposed solution is inefficient. The attention of the human operator becomes focused in a single wheel. In addition, the proposed solution cannot assure that the commands accomplish the physical constrains of the vehicle, resulting in slippage or even in clashes. This work proposes a solution that consists in the control of the vehicle looking at the position of its center of mass and its heading in the world frame. The solution is implemented using a rotational disk to control the vehicle heading and a common analogue joystick to control the vector speed of the center of the mass of the vehicle. The number of degrees of freedom reduces to three, i.e., two angles (vehicle heading and the orientation of the vector speed) and a scalar (the magnitude of the speed vector). This is possible using a kinematic model based on the vehicle Instantaneous

  9. Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Pedro; Vale, Alberto [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ventura, Rodrigo [Institute for Systems and Robotics, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2015-07-01

    The Cask and Plug Remote Handling System (CPRHS) and the respective Cask Transfer System (CTS) are designed to transport activated components between the reactor and the hot cell buildings of ITER during maintenance operations. In nominal operation, the CPRHS/CTS shall operate autonomously under human supervision. However, in some unexpected situations, the automatic mode must be overridden and the vehicle must be remotely guided by a human operator due to the harsh conditions of the environment. The CPRHS/CTS is a rhombic-like vehicle with two independent steerable and drivable wheels along its longitudinal axis, giving it omni-directional capabilities. During manual guidance, the human operator has to deal with four degrees of freedom, namely the orientations and speeds of two wheels. This work proposes a Human Machine Interface (HMI) to manage the degrees of freedom and to remotely guide the CPRHS/CTS in ITER taking the most advantages of rhombic like capabilities. Previous work was done to drive each wheel independently, i.e., control the orientation and speed of each wheel independently. The results have shown that the proposed solution is inefficient. The attention of the human operator becomes focused in a single wheel. In addition, the proposed solution cannot assure that the commands accomplish the physical constrains of the vehicle, resulting in slippage or even in clashes. This work proposes a solution that consists in the control of the vehicle looking at the position of its center of mass and its heading in the world frame. The solution is implemented using a rotational disk to control the vehicle heading and a common analogue joystick to control the vector speed of the center of the mass of the vehicle. The number of degrees of freedom reduces to three, i.e., two angles (vehicle heading and the orientation of the vector speed) and a scalar (the magnitude of the speed vector). This is possible using a kinematic model based on the vehicle Instantaneous

  10. Materials issues in cask development

    International Nuclear Information System (INIS)

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

    1987-01-01

    This paper identifies potential new materials as a function of their use in the cask. To the extent that identified materials are not yet qualified for their intended application, this paper identifies probable technical issues and development efforts that may be required to qualify the materials for use in transportation casks. 1 tab

  11. Transportation accident response of a high-capacity truck cask for spent fuel

    International Nuclear Information System (INIS)

    O'Connell, W.J.; Glaser, R.E.; Johnson, G.L.; Perfect, S.A.; McGuinn, E.J.; Lake, W.H.

    1995-11-01

    Two of the primary goals of this study were (i) to check the structural and thermal performance of the GA-4 cask in a broad range of accidents and (ii) to carry out a severe-accidents analysis as had been addressed in the Modal Study but now using a specific recent cask design and using current-generation computer models and capabilities. At the same time, it was desired to compare the accident performance of the Ga-4 cask to that of the generic truck cask analyzed in the Modal Study. The same range of impact and fire accidents developed in the Modal Study was adopted for this study. The accident-description data base of the Modal Study categorizes accidents into types of collisions with mobile or fixed objects, non-collision accidents, and fires. The mechanical modes of damage may be via crushing, impact, or puncture. The fire occurrences in the Modal Study data are based on truck accident statistics. The fire types are taken to be pool fires of petroleum products from fuel tanks and/or cargoes

  12. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of the document includes the LWT cask with fuel baskets; impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. 75 figs., 48 tabs

  13. Cask development, testing, and licensing

    International Nuclear Information System (INIS)

    Quinn, G.J.; Haelsig, R.T.; Warrant, M.M.

    1986-01-01

    The NuPac 125-B Rail Cask was developed to provide a safe means of transporting the damaged core of Three Mile Island Unit 2 from the TMI site at Middletown, PA, to the Idaho National Engineering laboratory (INEL) at Idaho Falls, ID. The development of the NuPac 125-B Rail Cask posed two engineering and technical management challenges; Licensing Strategy - The NuPac 125-B Rail Cask represented the first irradiated fuel rail cask developed within the United States in the past decade, a decade characterized by changing nuclear regulations, and Accelerated Schedule - The TMI-2 defueling schedule demanded a cask development schedule one-third as long as normally required. These challenges governed the overall development and licensing process for the cask. First, a high degree of conservation was incorporated into the design to allow quick, simplified demonstrations of adequacy to regulatory staff. Second, redundant design techniques were employed in all areas of uncertainty. The testing program eliminated performance uncertainties and validated predictions and predictive models. Drop tests of a quarter-scale model of the cask were conducted, and results were correlated with analytic predictions to verify structural and mechanical performance of the cask. Full-scale tests of the canisters were conducted to verify structural behavior of canister internals which provide criticality control. This paper describes the testing program for the NuPac 125-B Rail Cask, presents results therefrom, and correlates findings with Regulation 10 CFR 71 of the U.S. Nuclear Regulatory Commission

  14. 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.)

  15. 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

  16. Transport package response to severe thermal events, part 2: legal weight truck cask

    International Nuclear Information System (INIS)

    Greiner, M.; Faulkner, R.J.; Jin, Y.Y.

    1998-01-01

    The response of intact and damaged versions of the GA-4 Legal Weight Truck Cask to a range of severe thermal events is simulated using finite element computer analysis. The minimum fire durations that cause the containment seals and fuel cladding to reach their respective temperature limits are evaluated for a range of hydrocarbon fire temperatures. Containment seals reach their temperature limit in shorter duration fires as compared to the cladding, for both an undamaged package and a cask whose impact limiter is destroyed moments before the fire begins. However, if the neutron shield is destroyed, the cladding reaches its limit first in high temperature fires. A margin of safety exists between the conditions of the IAEA regulatory fire test and all of the performance envelopes calculated in this work. (author)

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

    International Nuclear Information System (INIS)

    Bauer, Thomas L.; Krause, Michael G.

    1986-01-01

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

  18. Crash test of a nuclear spent fuel cask and truck transport system

    International Nuclear Information System (INIS)

    Huerta, M.; Yoshimura, R.H.

    1978-01-01

    Sandia Laboratories has conducted a 96 kph (60 mph) full scale truck impact test for ERDA's Environmental Control Technology Division. Rockets propelled a 20, 500-kg (22-ton) cask mounted on its shipping trailer, coupled to a conventional cab-over tractor, into a massive, heavily reinforced concrete target. This summary report describes and compares the results of the computer analysis, scale model, and full scale tests

  19. 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

  20. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

  1. A preliminary investigation of the applicability of surface complexation modeling to the understanding of transportation cask weeping

    International Nuclear Information System (INIS)

    Granstaff, V.E.; Chambers, W.B.; Doughty, D.H.

    1994-01-01

    A new application for surface complexation modeling is described. These models, which describe chemical equilibria among aqueous and adsorbed species, have typically been used for predicting groundwater transport of contaminants by modeling the natural adsorbents as various metal oxides. Our experiments suggest that this type of modeling can also explain stainless steel surface contamination and decontamination mechanisms. Stainless steel transportation casks, when submerged in a spent fuel storage pool at nuclear power stations, can become contaminated with radionuclides such as 137 Cs, 134 Cs, and 60 Co. Subsequent release or desorption of these contaminants under varying environmental conditions occasionally results in the phenomenon known as open-quotes cask weeping.close quotes We have postulated that contaminants in the storage pool adsorb onto the hydrous metal oxide surface of the passivated stainless steel and are subsequently released (by conversion from a fixed to a removable form) during transportation, due to varying environmental factors, such as humidity, road salt, dirt, and acid rain. It is well known that 304 stainless steel has a chromium enriched passive surface layer; thus its adsorption behavior should be similar to that of a mixed chromium/iron oxide. To help us interpret our studies of reversible binding of dissolved metals on stainless steel surfaces, we have studied the adsorption of Co +2 on Cr 2 O 3 . The data are interpreted using electrostatic surface complexation models. The FITEQL computer program was used to obtain the model binding constants and site densities from the experimental data. The MINTEQA2 computer speciation model was used, with the fitted constants, in an attempt to validate this approach

  2. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  3. 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

  4. 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

  5. Radiological source terms resulting from sabotage to transportation casks: Final report

    International Nuclear Information System (INIS)

    Miller, N.E.; Fentiman, A.W.; Kuhlman, M.R.; Ebersole, H.N.; Trott, B.D.; Orban, J.E.

    1986-11-01

    The Nuclear Regulatory Commission (NRC) promulgated a rule, 10 CFR 73.37, which established requirements for safeguarding shipments of spent fuel to reduce the risk from acts of sabotage of highly radioactive materials. After the rule became effective, experimental programs conducted by Battelle for the NRC and by Sandia for the DOE showed the consequences of an attack using explosives on a shipment of PWR spent fuel were significantly less than had been indicated by earlier analytical studies. As a result, NRC is considering modifying the safeguards requirements. In support of NRC's efforts to modify the rule, Battelle has conducted additional experimental studies to evaluate the consequences of attacks on shipments of high-temperature gas-cooled reactor (HTGR) spent fuel, nonpower reactor (NPR) spent fuel, and vitrified high-level waste (HLW). Model casks containing surrogates of the spent fuels or high-level waste were penetrated by the jet from a precision shaped charge. Air samples collected after each test were used to estimate the quantities of respirable material released after the cask was penetrated. Results of the tests were scaled by specially developed scaling factors to estimate the releases that may occur from attacks on full-sized shipments of the materials. It was concluded that the sabotage of shipments of HTGR spent fuel, NPR spent fuel, or vitrified HLW should have no greater consequences than those predicted for shipments of PWR spent fuel

  6. 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

  7. 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.

  8. Determination of uncertainties in the calculation of dose rates at transport and storage casks; Unsicherheiten bei der Berechnung von Dosisleistungen an Transport- und Lagerbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Schloemer, Luc Laurent Alexander

    2014-12-17

    The compliance with the dose rate limits for transport and storage casks (TLB) for spent nuclear fuel from pressurised water reactors can be proved by calculation. This includes the determination of the radioactive sources and the shielding-capability of the cask. In this thesis the entire computational chain, which extends from the determination of the source terms to the final Monte-Carlo-transport-calculation is analysed and the arising uncertainties are quantified not only by benchmarks but also by variational calculi. The background of these analyses is that the comparison with measured dose rates at different TLBs shows an overestimation by the values calculated. Regarding the studies performed, the overestimation can be mainly explained by the detector characteristics for the measurement of the neutron dose rate and additionally in case of the gamma dose rates by the energy group structure, which the calculation is based on. It turns out that the consideration of the uncertainties occurring along the computational chain can lead to even greater overestimation. Concerning the dose rate calculation at cask loadings with spent uranium fuel assemblies an uncertainty of (({sup +21}{sub -28}) ±2) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are estimated. For mixed-loadings with spent uranium and MOX fuel assemblies an uncertainty of ({sup +24±3}{sub -27±2}) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are quantified. The results show that the computational chain has not to be modified, because the calculations performed lead to conservative dose rate predictions, even if high uncertainties at neutron dose rate measurements arise. Thus at first the uncertainties of the neutron dose rate measurement have to be decreased to enable a reduction of the overestimation of the calculated dose rate afterwards. In the present thesis

  9. 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

  10. 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

  11. 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.

  12. Prototypic fabrication of TRIGA irradiated fuel shipping casks

    International Nuclear Information System (INIS)

    Kim, B.K.; Lee, Y.W.; Whang, C.K.; Lee, J.B.

    1980-01-01

    This is the safety analysis report on the prototypic fabrication of ''TRIGA Irradiated Fuel Shipping Cask'' conducted by KAERI in 1980. The results of the evaluation show that the shipping cask is in compliance with the applicable regulation for the normal conditions of transport as well as hypothetical accident conditions. The prototypic fabrication of the shipping cask (type B) was carried out for the first time in Korea after getting technical experience from fabrication of the ''TRIGA Spent Fuel Shipping Cask'' and ''the KO-RI Unit 1 surveillance capsule shipping cask'' in 1979. This report contains structural evaluation, thermal evaluation, shielding, criticality, quality assurance, and handling procedures of the shipping cask

  13. A preliminary assessment of system cost impacts of using transportable storage casks and other shippable metal casks in the utility/DOE spent fuel management system

    International Nuclear Information System (INIS)

    Johnson, E.R.

    1988-01-01

    In view of the foregoing, a study was conducted by E.R. Johnson Associates, Inc. and H and R Technical Associates, Inc. to determine the prospective viability of the use of TSCs and shippable SOCs in the combined utility/DOE system. This study considered costs, ALARA considerations and the logistics of the use and delivery of casks to the DOE system by utilities. It was intended that this study would result in a technical and cost resource base that could be used for evaluating various strategies and scenarios for deploying TSCs or SOCs in the combined utility/DOE spent fuel management system with respect to the prospective economic advantage that could be realized

  14. 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

  15. 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

  16. Condensate-polisher resin-leakage quantification and resin-transport studies

    International Nuclear Information System (INIS)

    Stauffer, C.C.; Doss, P.L.

    1983-04-01

    The objectives of this program were to: (1) determine the extent of resin leakage from current generation condensate polisher systems, both deep bed and powdered resin design, during cut-in, steady-state and flow transient operation, (2) analyze moisture separator drains and other secondary system samples for resin fragments and (3) document the level of organics in the secondary system. Resin leakage samples were obtained from nine-power stations that have either recirculating steam generators or once through steam generators. Secondary system samples were obtained from steam generator feedwater, recirculating steam generator blowdown and moisture separator drains. Analysis included ultraviolet light examination, SEM/EDX, resin quantification and infrared analysis. Data obtained from the various plants were compared and factors affecting resin leakage were summarized

  17. Force and moment reconstruction for a nuclear transportation cask using sum of weighted accelerations and deconvolution theory

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Bateman, V.; Carne, T.G.; Gregory, D.L.; Attaway, S.W.; Bronowski, D.R.

    1989-01-01

    A 9-m drop test was conducted of a 1/3-scale-model spent fuel cask onto an unyielding target. The structural response of the impact limiters and attachments was evaluated. A mass model of the cask body, with steel-sheathed redwood and balsa impact limiters, was tested in a 10-degree slapdown orientation. One end of the cask impact the target before the other end, with higher deceleration forces resulting from the second impact. The information desired from this test is the deformation of the two impact limiters on either end of the cask as a function of the applied force. The content in this paper will only discuss a summary of the applied force calculations. Additional details about the force and moment reconstruction methods and analysis results and test and hardware are provided elsewhere. Two new force reconstruction techniques were applied to the slapdown test data: the sum of weighted accelerations technique (SWAT) and deconvolution (DECON). The rigid-body acceleration is then multiplied by the cask mass to obtain an estimate of the applied force. The frequency content of this force is restricted to the cut-off frequency of the digital filter, typically about one-half of the lowest elastic mode of the cask. The new force reconstruction techniques demonstrate the potential for a better estimate of forces acting on the cask during the impact than the conventional method. The new force reconstruction techniques use the cask structure as a generalized force transducer. With these techniques, the elastic vibration response of the cask is eliminated from the acceleration data. The main advantages of the force reconstruction techniques are the extension of the frequency bandwidth (due to the elimination of the elastic modal response in that bandwidth) and the preservation of the force rise time

  18. A CASKCOM: A cask life cycle cost model

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    CASKCOM (cask cost model) is a computerized model which calculates the life cycle costs (LCC) associated with specific transportation cask designs and discounts those costs, if the user so chooses, to a net present value. The model has been used to help analyze and compare the life cycle economics of burnup credit and nonburnup credit cask designs being considered as conditions for a new generation of spent fuel transportation casks. CASKCOM is parametric in the sense that its input data can be easily changed in order to analyze and compare the life cycle cost implications arising from alternative assumptions. The input data themselves are organized into two main groupings. The first grouping comprises a set of data which is independent of cask design. This first grouping does not change from the analysis of one cask design to another. The second grouping of data is specific to each individual cask design. This second grouping thus changes each time a new cask design is analyzed

  19. Numerical study of thermal test of a cask of transportation for radioactive material; Estudo numérico do ensaio térmico de um casco de transporte para material radioativo

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Tiago A.S.; Santos, André A.C. dos; Vidal, Guilherme A.M.; Silva Junior, Geraldo E., E-mail: tiago.vieira.eng@gmail.com, E-mail: gvidal.ufmg@gmail.com, E-mail: aacs@cdtn.br, E-mail: geraldo.esilva@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    In this study numerical simulations of a transport cask for radioactive material were made and the numerical results were compared with experimental results of tests carried out in two different opportunities. A mesh study was also made regarding the previously designed geometry of the same cask, in order to evaluate its impact in relation to the stability of numerical results for this type of problem. The comparison of the numerical and experimental results allowed to evaluate the need to plan and carry out a new test in order to validate the CFD codes used in the numerical simulations.

  20. Developing a brittle fracture acceptance criterion for transport casks for adoption by the International Atomic Energy Agency(IAEA)

    International Nuclear Information System (INIS)

    Sorenson, K.B.; Falci, F.P.

    1991-01-01

    With increasing nuclear material transportation demands, new generation casks have included designs which propose the use of structural materials other than austenitic stainless steel. Motivation for using alternate materials include potentials for lower cost, easier fabrication, no welding and less weight (higher payload). Examples of candidate materials for structural components include ferritic steels and ductile cast iron for the containment boundary and borated stainless steel for the basket. The primary technical issue which separates these candidate materials from austenitic stainless steel is that they may, under certain combinations of mechanical and environmental loadings, fail in a brittle fracture mode. There is a clear need to develop a consensus criterion that will encompass a broad range of materials and provide a consistent method for evaluating candidate materials with respect to brittle fracture. Adopting such a criterion through the auspices of the IAEA has distinct advantages. An international consensus would provide assurance to regulators and the public that the most appropriate method is used, and countries developing packages using candidate materials could work with a single criterion without concern that another country is setting precedent using a different criterion. The development of such a criterion has been in progress for two and one-half years. There are two major efforts which are being pursued in parallel. There is the technical effort to write a draft document and the formal effort to facilitate adoption of a criterion by the IAEA. 7 refs., 1 fig

  1. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  2. FUEL CASK IMPACT LIMITER VULNERABILITIES

    International Nuclear Information System (INIS)

    Leduc, D.; England, J.; Rothermel, R.

    2009-01-01

    Cylindrical fuel casks often have impact limiters surrounding just the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and maintaining lower peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) impacts. Large casks are often certified by analysis only because of the costs associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the impact limiter attachment system. Assumptions in the original Safety Analysis for Packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs

  3. Dry cask handling system for shipping nuclear fuel

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

    A nuclear facility is described for improved handling of a shipping cask for nuclear fuel. After being brought into the building, the cask is lowered into a tank mounted on a transporter, which then carries the tank into a position under an auxiliary well to which it is sealed. Fuel can then be loaded into or unloaded from the cask via the auxiliary well which is flooded. Throughout the procedure, the cask surface remains dry. (U.S.)

  4. TITAN Legal Weight Truck cask preliminary design report

    International Nuclear Information System (INIS)

    1990-04-01

    The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of this document includes the LWT cask with fuel baskets, impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. The results of the tradeoff studies and evaluations that were performed during the preliminary design are presented in Appendix A to this report. 51 figs., 17 tabs

  5. 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

  6. 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

  7. Inspection of NFT-type cask fabrication

    International Nuclear Information System (INIS)

    Takani, M.; Umegaki, O.

    1998-01-01

    NFT-type cask has been developed to transport the high burn-up spent fuel from Japanese nuclear power stations to the reprocessing plant of Japan Nuclear Fuel Limited which is under construction in Rokkasho-mura, Aomori prefecture. NFT placed orders of 53 casks to 5 fabricators in Japan and overseas, and these casks have been fabricated since 1994. There are two types of NFT-type casks for PWR spent fuel and four types of NFT-type cask for BWR spent fuel. These are designed in consideration of the number of spent fuels accommodated into each type of casks and the handling conditions at domestic nuclear power stations. According to Japanese notification, it is required to be confirmed by competent authority that casks are manufactured in accordance with approved designs. Furthermore, additional tests are performed such as through-gauge test for basket and pressure test on the shielding material space to ensure the performance of cask by NFT other than items inspected by the competent authority. In order to enhance maintainability of casks, replacement parts such as bolts and valves are shared as much as possible. (authors)

  8. Nuclear criticality safety: general. 5. Reactivity Effect of Burnable Absorbers in Burnup Credit for the CASTOR X/32S Storage and Transport Cask

    International Nuclear Information System (INIS)

    Rombough, Charles T.; Lancaster, Dale B.; Diersch, Rudolf; Spilker, Harry

    2001-01-01

    When considering burnup credit in the licensing of storage and transportation casks, a significant effect is whether or not the burned fuel was depleted with burnable absorbers present. This paper presents the results of detailed calculations to quantitatively determine the burnable absorber effect for the CASTOR X/32S transport cask, which assumes burnup of the fuel in the criticality analysis. A radial view of the CASTOR X/32S cask is shown in Fig. 1. This is the actual plot of the geometry as modeled in KENO V.a. Note that there are no water-filled flux traps and the assemblies are tightly packed. This reduces the overall dimensions of the cask for a given number of fuel assemblies. Reactivity is held down by borated aluminum plates between the fuel assemblies and by placing absorber rod modules (ARMs) in the guide tubes of selected assemblies. If burnup of the fuel is not considered and the initial enrichment is 5.0 wt% 235 U, then 28 of the 32 fuel assemblies must contain an ARM to maintain a k eff 3 ; 4. moderator temperature of 604 K; 5. cooling time of 9.5 yr; 6. specific power of 60 W/g of U metal; 7. conservative axial and radial burnup shape distribution; 8. Westinghouse BP material containing 12.5 wt% B 4 C. Using the model described earlier, calculations were performed with varying numbers of BP fingers inserted for different exposure times. The results are shown in Tables I and II. The 1 s statistical error in these results is σ equals ±0.05%. Note that the BP finger and exposure effects decrease with fuel burnup and the effect is smaller when the cask contains ARMs. Conservatively combining the results from Tables I and II and interpolating, we can equate fewer BP fingers with longer BP exposure time as shown in Table III. The Table III results were checked by running the actual cases (for example, 20 BP fingers for 24 GWd/tonne exposure) to verify that the k eff 's for the cask were always less than the base-case values. These results can also be

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

    International Nuclear Information System (INIS)

    1986-07-01

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

  10. Operations experience with the NAC-1 legal weight truck cask

    International Nuclear Information System (INIS)

    Viebrock, J.M.; Hoffman, C.C.

    1978-01-01

    The first three years of operation of Nuclear Assurance Corporation's (NAC) four (4) NAC-1 Casks have demonstrated that shipments of spent fuel, fuel rods and other highly irradiated reactor components can be moved routinely by legal weight truck transport. Shipments of these materials have involved some 800,000 miles of highway travel and cask handling at some fifteen different nuclear facilities. This paper presents details on NAC's operations experience with these casks including cask description, cask handling (loading and unloading), pre-shipment testing, facility turnaround and transit times, operator exposure, transport vehicles and shipper/carrier/cask owner responsibilities, actual experience with regard to facility interfacing requirements and operational procedures. Cask and equipment utilization is discussed together with the methods used to control operation costs and to improve the economics of truck transport

  11. 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

  12. THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-06-01

    The Global Threat Reduction Initiative (GTRI) has many experiments yet to be irradiated in support of the High Performance Research Reactor fuels development program. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for post irradiation examination. To date, the General Electric (GE)-2000 cask has been used to transport GTRI experiments between these facilities. However, the availability of the GE-2000 cask to support future GTRI experiments is at risk. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger GTRI experiments. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled experiments. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask.

  13. Impact analysis of shipping casks

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1989-01-01

    Shipping casks are being used in the United States Department of Energy to transport irradiated experiments, reactor fuel, radioactive waste, etc. One of the critical requirements in shipping cask analysis is the necessity to withstand severe impact environments. It is still conventional to develop the design and to verify the design requirements by hand calculations. Full three dimensional computations of impact scenarios have been performed but they are too expensive and time consuming for design purposes. Typically, on the order of more than an hour of CRAY time is required for a detailed, three dimensional analysis. The paper describes how simpler two- and three-dimensional models can be used to provide an intermediate level of detail between full three dimensional finite element calculations and hand calculations. The regulation that is examined here is: 10 CFR-71.73 hypothetical accident conditions, free drop. Free drop for an accident condition of a Class I package (approximate weight of 22,000 lb) is defined as a 30 foot drop onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. Three free drop scenarios are analyzed to assess the integrity of the cask when subjected to large bending and axial stresses. These three drop scenarios are: (1) a thirty foot axial drop on either end, (2) a thirty foot oblique angle drop with the cask having several different orientations from the vertical with impact on the top end cask corner, and (3) a thirty foot side drop with simultaneous impact on the strength of the various components that comprise the cask. The predicted levels of deformation and stresses in the cask will be used to assess the potential damage level. 5 refs., 5 figs., 1 tab

  14. Demonstration of criticality safety for the modified TN-REG and TN-BRP transport/storage casks

    International Nuclear Information System (INIS)

    Parks, C.V.; Fox, P.B.

    1989-01-01

    An inability to model the structural performance of borated steel baskets under accident conditions forced the specially designed TN-BRP and TN-REG casks to be modified for half-loaded shipments. This paper discusses the approach used to demonstrate that the half-loaded casks would remain safely subcritical even if no credit were taken for the borated basket. Normal and accident configurations were analyzed with the KENO V.a code. The strategy conceived and the analyses performed to demonstrate an acceptable margin of safety are discussed. 5 refs., 3 figs., 2 tabs

  15. Cask fleet operations study

    International Nuclear Information System (INIS)

    1988-01-01

    The Nuclear Waste Policy Act of 1982 assigned to the Department of Energy's (DOE) Office of Civilian Waste Management the responsibility for disposing of high-level waste and spent fuel. A significant part of that responsibility involves transporting nuclear waste materials within the federal waste management system; that is, from the waste generator to the repository. The lead responsibility for transportation operations has been assigned to Oak Ridge Operations, with Oak Ridge National Laboratory (ORNL) providing technical support through the Transportation Operations Support Task Group. One of the ORNL support activities involves assessing what facilities, equipment and services are required to assure that an acceptable, cost-effective and safe transportation operations system can be designed, operated and maintained. This study reviews, surveys and assesses the experience of Nuclear Assurance Corporation (NAC) in operating a fleet of spent-fuel shipping casks to aid in developing the spent-fuel transportation system

  16. Functions of the cask maintenance facility: A white paper

    International Nuclear Information System (INIS)

    1987-01-01

    The shipping cask systems are the mobile components of the transportation system, designed to safely transport spent nuclear fuel between different facilities under both normal and accident conditions. The cask system will consist of the heavily shielded cask, the cask transport vehicle (truck trailer or railcar), and any associated ancillary equipment (covers, impact limiters, lifting devices, etc.). The cask and certain parts of the cask system must be operated within the limits imposed by a certificate of compliance (COC) granted by the Nuclear Regulatory Commission (NRC). Each cask system must transport spent fuel safely during the life of the system. To maintain the operational effectiveness and safety of the cask systems, a cask maintenance facility (CMF) will be included as an integral part of the transportation system. The planning activity of the transportation system and the design effort of the CMF require that the functions to be performed by the CMF be explicitly defined. The purpose of this paper is to (1) define the potential transportation system functions to be performed at the CMF; (2) examine the impact of this functional definition on the overall transportation system; (3) identify any unresolved issues concerning the interaction of the CMF with other elements of the transportation system; and (4) make recommendations to resolve any unresolved issues so that decisions can be made early in the transportation system planning process

  17. 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

  18. 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

  19. 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.)

  20. 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

  1. 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)

  2. Design review report FFTF interim storage cask

    International Nuclear Information System (INIS)

    Scott, P.L.

    1995-01-01

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location

  3. 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

  4. 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

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

    SMITH, R.J.

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, R.J.

    1998-11-17

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

  9. Study of the temperature distribution on welded thin plates of duplex steel to be used for the external clad of a cask for transportation of radiopharmaceuticals products

    International Nuclear Information System (INIS)

    Betini, Evandro G.; Ceoni, Francisco C.; Mucsi, Cristiano S.; Politano, Rodolfo; Rossi, Jesualdo L.; Orlando, Marcos T.D.

    2015-01-01

    The clad material for a proprietary transport device for radiopharmaceutical products is the main focus of the present work. The production of 99 Mo- 99m Tc transport cask requires a receptacle or cask where the UNS S32304 duplex steel sheet has shown that it meets high demands as the required mechanical strength and the spread of impact or shock waves mitigation. This work reports the experimental efforts in recording the thermal distribution on autogenous thin plates of UNS S32304 steel during welding. The UNS S32304 duplex steel is the most probable candidate for the external clad of the containment package for the transport of radioactive substances so it is highly relevant the understanding of all its physical parameters and its behavior under the thermal cycle imposed by a welding process. For the welding of the UNS S32304 autogenous plates the GTAW (gas tungsten arc welding) process was used with a pure argon arc protection atmosphere in order to simulate a butt joint weld on a thin duplex steel plate without filler metal. The thermal cycles were recorded by means of K-type thermocouples embedded by electrical spot welding near the weld region and connected to a multi-channel data acquisition system. The obtained results validate the reliability of the experimental apparatus for the future complete analysis of the welding experiment and further comparison to numerical analysis. (author)

  10. 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)

  11. Natural ventilation of a generic cask under a transport hood - CFD and analytical modelling

    Energy Technology Data Exchange (ETDEWEB)

    Powell, D.; Davies, G.; Tso, C.F. [Arup, London (United Kingdom)

    2004-07-01

    In comparison with finite element simulation for structural and thermal behaviour, the use of computational fluid dynamics technique (hereafter CFD) to analyse, predict and design air and heat flow in package design is relatively novel. Arup has been using CFD techniques to investigate fluid and heat flow, and to use it as a tool to design fluid and heat flow across a broad spectrum of industries for over fifteen years. In order demonstrate the power of the technique and its benefits, the airflow and heat flow characteristics around a transport package during transit under a transport hood has been evaluated using the CFD technique. This paper presents the scenario, the model, the analysis technique and the results of this analysis. Comparison with test results is probably the best way to validate a CFD analysis. In the absence of test results, the analysis was verified by comparison with hand calculation solutions. The scenario as it stands is too complex and hand calculation solution cannot describe the scenario sufficiently. However, hand calculation solutions could be derived for simplified version of the scenario against which CFD analysis of the simplified scenario can be compared. The second half of this paper describes the verification out.

  12. Natural ventilation of a generic cask under a transport hood - CFD and analytical modelling

    International Nuclear Information System (INIS)

    Powell, D.; Davies, G.; Tso, C.F.

    2004-01-01

    In comparison with finite element simulation for structural and thermal behaviour, the use of computational fluid dynamics technique (hereafter CFD) to analyse, predict and design air and heat flow in package design is relatively novel. Arup has been using CFD techniques to investigate fluid and heat flow, and to use it as a tool to design fluid and heat flow across a broad spectrum of industries for over fifteen years. In order demonstrate the power of the technique and its benefits, the airflow and heat flow characteristics around a transport package during transit under a transport hood has been evaluated using the CFD technique. This paper presents the scenario, the model, the analysis technique and the results of this analysis. Comparison with test results is probably the best way to validate a CFD analysis. In the absence of test results, the analysis was verified by comparison with hand calculation solutions. The scenario as it stands is too complex and hand calculation solution cannot describe the scenario sufficiently. However, hand calculation solutions could be derived for simplified version of the scenario against which CFD analysis of the simplified scenario can be compared. The second half of this paper describes the verification out

  13. CASTOR {sup registered} 1000/19. Transport and storage cask for the disposal of spent fuel from the nuclear power plant Temelin (Czech Republic); CASTOR {sup registered} 1000/19. Transport- und Lagerbehaelter zur Entsorgung abgebrannter Brennelemente aus dem Kernkraftwerk Temelin in Tschechien

    Energy Technology Data Exchange (ETDEWEB)

    Fopp, Stefan; Kuehne, Bernhard; Schroeder, Jens [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2011-07-01

    The transport and storage cask CASTOR {sup registered} 1000/19 was designed for a dry interim storage of 19 spent fuel elements of WWER-1000 reactors. The project performed by GNS mbH included design, manufacture, licensing and delivery of 36 casks. The specific requirements for NPP Temelin concern loading and dispatch of the casks to be performed during the outage period of the reactor, thus high reliability and functionality of the casks and the manipulation equipment. The paper describes the mechanical design of the cask, stress analyses for a hypothetical fall accident from 9 m height, performed using the FEM program ANSYS and LS-DYNA. The 3D simulation models are based on conservative material characteristics and upper-bound boundary conditions. The safety analysis was performed using qualified software programs validated by the Czech authorities.

  14. Development on application of ultrasonic sealing techniques to plutonium transportation cask

    International Nuclear Information System (INIS)

    Hayakawa, Tsuyoshi; Akiba, Mitsunori; D'Agraives, B.C.

    1994-01-01

    In a cooperation research between Power Reactor and Nuclear Fuel Development Corporation and Commission of the European Communities, Joint Research Centre, Ispra establishment, application of ultrasonic sealing techniques to a plutonium container is developed in Ispra. The seal is derived from the sealing-bolt technology currently in use at the BNFL site of sellafield (UK) for the safeguarding of underwater spent fuel storage containers called MEBs. In this technique, one of the normal bolts closing the lid of the container is replaced by a special ultrasonically verifiable sealing-bolt. In the application to the plutonium container, it is proposed to attach a clamping seal which has the same internal configuration as a MEB sealing-bolt but is fastened with a 'one-way' mechanism to one of the protruding pins of the container. Similarly the seal is provided with an identity and integrity features. The uniqueness of the identity, as well as the integrity can be checked on the spot by an inspector carrying a reading equipment. Thus, in a few minutes, one identifies the seal and knows whether its integrity is intact, which tells that the container has not been opened or attempted to open illeagally. By application of the seal to the plutonium container, the containment/surveillance during the transportation will be upgraded. (author)

  15. Evaluation method of gas leakage rate from transportation casks of radioactive materials (gas leakage rates from scratches on O-ring surface)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Li Ninghua; Asano, Ryoji; Kawa, Tsunemichi

    2004-01-01

    A sealing function is essential for transportation and/or storage casks of radioactive materials under both normal and accidental operating conditions in order to prevent radioactive materials from being released into the environment. In the safety analysis report, the release rate of radioactive materials into the environment is evaluated using the correlations specified in the ANSI N14.5, 1987. The purposes of the work are to reveal the underlying problems on the correlations specified in the ANSI N14.5 related to gas leakage rates from a scratch on O-ring surface and from multi-leak paths, to offer a data base to study the evaluation method of the leakage rate and to propose the evaluation method. In this paper, the following insights were obtained and clarified: 1. If a characteristic value of a leak path is defined as D 4 /a ('D' is the diameter and 'a' is the length), a scratch on the O-ring surface can be evaluated as a circular tube. 2. It is proper to use the width of O-ring groove on the flange as the leak path length for elastomer O-rings. 3. Gas leakage rates from multi leak paths of the transportation cask can be evaluated in the same manner as a single leak path if an effective D4/a is introduced. (author)

  16. 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)

  17. DOE procurement activities for spent fuel shipping casks

    International Nuclear Information System (INIS)

    Callaghan, E.F.; Lake, W.H.

    1988-01-01

    The DOE cask development program satisfies the requirements of the NWPA by providing safe efficient casks on a timely schedule. The casks are certified by the NRC in compliance with the 1987 amendment to NWPA. Private industry is used to the maximum extent. DOE encourages use of present cask technology, but does not hesitate to advance the state-of-the-art to improve efficiency in transport operations, provided that safety is not compromised. DOE supports the contractor's efforts to advance the state-of-the-art by maintaining a technical development effort that responds to the common needs of all the contractors. DOE and the cask contractors develop comprehensive and well integrated programs of test and analysis for cask certification. Finally, the DOE monitors the cask development program within a system that fosters early identification of improvement opportunities as well as potential problems, and is sufficiently flexible to respond quickly yet rationally to assure a fully successful program

  18. Cask manufacturing methods and quality assurance problems

    International Nuclear Information System (INIS)

    Riddle, J.H.; Cross, H.D.; Trujillo, A.A.; Pope, R.B.; Rack, H.J.

    1980-01-01

    This paper presents the results of a study performed to identify and evaluate cask manufacturing methods and materials which are presently or could be readily available in the United States. The study includes a comparison of the economic and technical advantages to be gained from manufacturing casks specifically for application to long-cooled fuel (5 years or greater) and waste materials. The conclusions that have been drawn from this study, which is still in progress are as follows: if spent fuel payloads are in fact cooled for five or more years prior to transportation there is a significant advantage to be gained from cask designs which are tailored to this longer cooling time; the most economic cask design is that which utilizes a lead gamma shield; the all-steel, single wall casks show substantial potential benefits resulting from ready availability and the cost and inspection advantages that result from the design simplicity; the future use of depleted uranium shielded casks in the United States is expected to be low as a result of the high cost and lack of manufacturing facilities; of the three types of neutron absorber systems evluated, the borated cast-in-place silicone rubber shows significant technical and cost advantages; and design and fabrication code requirements and regulatory requirements yet to be developed in the United States are likely to have a profound effect on the next generation of transportation casks

  19. Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

    International Nuclear Information System (INIS)

    Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

    2003-01-01

    The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

  20. 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

  1. 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)

  2. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis

    International Nuclear Information System (INIS)

    Neumann, Martin

    2009-01-01

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  3. 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

  4. ALARA Principle Application for Loading Spent Nuclear Fuel Assemblies from Nuclear Research Reactor WR-S Mergal-Bucharest Romania into Transportation Casks

    International Nuclear Information System (INIS)

    Dragusin, M.

    2009-01-01

    Safety implementation of Spent Nuclear Fuels Assemblies (SNFA) handling procedures at the WR-S reactor site is ensured by technical perfection and reliability of equipment, monitoring of its condition, qualification and discipline of personnel as well as organization and execution of work complied with requirements of regulatory documents, process procedures, guidance and manuals. The personnel training for execution loading of SNF FAs is other important aspect for radiation protection and safely activities. Estimations carried out using Micro Shield software show that maximal dose rate upon working site when loading four FAs into basket of cask will not exceed 1.7 and 956;Sv/h, excluding natural radiation. Radiation Safety Analyses estimates for loading 70 SNFA in 18 transportation casks are: maximal individual dose: 4274.7 and 956;Sv, maximal expected collective dose persons: 17 031.2 man and 956;Sv. By application ALARA principle with technical and administrative measures the loading process developed in the following conditions: maximal individual dose: 68 and 956;Sv, the collective dose persons: 732 man and 956;Sv. The work will presented the technical measures and procedures applied in loading process.

  5. Characterization of dextran-grafted hydrophobic charge-induction resins: Structural properties, protein adsorption and transport.

    Science.gov (United States)

    Liu, Tao; Angelo, James M; Lin, Dong-Qiang; Lenhoff, Abraham M; Yao, Shan-Jing

    2017-09-29

    The structural and functional properties of a series of dextran-grafted and non-grafted hydrophobic charge-induction chromatographic (HCIC) agarose resins were characterized by macroscopic and microscopic techniques. The effects of dextran grafting and mobile phase conditions on the pore dimensions of the resins were investigated with inverse size exclusion chromatography (ISEC). A significantly lower pore radius (17.6nm) was found for dextran-grafted than non-grafted resins (29.5nm), but increased salt concentration would narrow the gap between the respective pore radii. Two proteins, human immunoglobulin G (hIgG) and bovine serum albumin (BSA), were used to examine the effect of protein characteristics. The results of adsorption isotherms showed that the dextran-grafted resin with high ligand density had substantially higher adsorption capacity and enhanced the salt-tolerance property for hIgG, but displayed a significantly smaller benefit for BSA adsorption. Confocal laser scanning microscopy (CLSM) showed that hIgG presented more diffuse and slower moving adsorption front compared to BSA during uptake into the resins because of the selective binding of multiple species from polyclonal IgG; polymer-grafting with high ligand density could enhance the rate of hIgG transport in the dextran-grafted resins without salt addition, but not for the case with high salt and BSA. The results indicate that microscopic analysis using ISEC and CLSM is useful to improve the mechanistic understanding of resin structure and of critical functional parameters involving protein adsorption and transport, which would guide the rational design of new resins and processes. Copyright © 2017. Published by Elsevier B.V.

  6. 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

  7. Applying consensus standards to cask development

    International Nuclear Information System (INIS)

    Leatham, J.; Abbott, D.G.; Warrant, M.M.

    1987-01-01

    The Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management is procuring cask systems for transporting commercial spent nuclear fuel and is encouraging development of innovative cask designs and materials to improve system efficiency. New designs and innovative materials require that consensus standards be established so that cask designers and regulators have criteria for determining acceptability. Recent DOE experience in certifying three spent fuel shipping casks, NUPAC-125B, TN-BRP, and TN-REG, is discussed. Certification of the NUPAC-125B was expedited because it was made of conventional American Society for Testing and Materials (ASTM) materials and complied with the American Society of Mechanical Engineers (ASME) Code and Nuclear Regulatory Commission Regulatory Guides. The TN-BRP and TN-REG cask designs are still being reviewed because baskets included in the casks are made of borated stainless steel, which has no ASTM Specification or ASME Code approval. The process of developing and approving consensus standards is discussed, including the role of ANSI and ANSI N14. Specific procedures for ASTM and ASME are described. A draft specification or standard must be prepared and then approved by the appropriate body. For new material applications to the ASME Code, an existing ASTM Specification is needed. These processes may require several years. The status of activities currently in progress to develop consensus standards for spent fuel casks is discussed, including (1) ASME NUPAC, and (2) ASTM Specifications for ductile cast iron and borated stainless steel

  8. The use of burnup credit for spent fuel cask design

    International Nuclear Information System (INIS)

    Lake, W.H.

    1993-01-01

    A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

  9. Interim and final storage casks

    International Nuclear Information System (INIS)

    Stumpfrock, L.; Kockelmann, H.

    2012-01-01

    The disposal of radioactive waste is a huge social challenge in Germany and all over the world. As is well known the search for a site for a final repository for high-level waste in Germany is not complete. Therefore, interim storage facilities for radioactive waste were built at plant sites in Germany. The waste is stored in these storage facilities in appropriate storage and transport casks until the transport in a final repository can be carried out. Licensing of the storage and transport casks aimed for use in the public space is done according to the traffic laws and for handling in the storage facility according to nuclear law. Taking into account the activity of the waste to be stored, different containers are in use, so that experience is available from the licensing and operation in interim storage facilities. The large volume of radioactive waste to be disposed of after the shut-down of power generation in nuclear power stations makes it necessary for large quantities of licensed storage and transport casks to be provided soon.

  10. Application of a fracture toughness analysis for ferritic steel components of transport/storage casks using an adapted EUROCODE 3 approach

    International Nuclear Information System (INIS)

    Sedlacek, G.; Dahl, W.; Langenberg, P.; Stranghoner, N.; Dreier, G.; Diersch, R.

    1998-01-01

    The choice of a structural steel material for a component of a transport and storage cask for RAM is strongly determined by the demand that the cask has to withstand a free fall from 9 m height without losing its integrity and leak tightness. In terms of fracture mechanics this means that instable crack growth must not occur even under the conditions of high amplitude dynamic loading at temperatures of -40 deg. C. In the course of harmonization of European design guidelines, the Eurocode 3 has been developed which contains a fracture mechanic based concept for the steel selection to avoid brittle fracture, called Annex C. This method combines fracture mechanics tools like the failure assessment diagram (CEGB-R6-procedure) with fracture mechanics life time assessment procedure for fatigue loaded structures. The required toughness in terms of the stress intensity factor K I is related to the T 27J Charpy transition temperature by means of a master curve and by a correlation between the fracture mechanics transition temperature T K100 and the Charpy transition temperature T 27J . Both relations have been proved to be valid for structural steels in the range of 235 to 960 MPa yield strength. Besides that a semi-probabilistic safety approach that takes account of the model inaccuracies by calibration of large scale tests has been applied to derive a safety element for a risk of failure of p f 10 -5 . The fracture mechanic concept of Eurocode 3 has been adopted to calculate critical failure lengths for lids made from ferritic steels of transport and storage casks. The safety requirements of Appendix VI of the IAEA Advisory Material have been taken into account. It has been shown that the adopted Eurocode 3, Annex C, method allows an economical calculation of critical failure length on a high level of safety. A failure probability of p f = 8 . 10 -7 is reached by applying lower bound estimates of fracture toughness and an additional additive safety factor ΔT a of 20 deg

  11. 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

  12. 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)

  13. 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

  14. Literature study of volatile radioiodine release from ion-exchange resins during transportation

    International Nuclear Information System (INIS)

    Wren, J.C.

    1991-02-01

    A transport package is currently being developed by Ontario Hydro to carry used filters and ion-exchange columns from the Pickering and Darlington Nuclear Generating Stations to the Bruce Nuclear Generating Station for disposal. The main reason that the transport package must be licensed is the possibility that volatile radionuclides being transported in the package might be released during transport accidents. Of particular concern is the iodine that might become volatile due to the degradation of the ion exchange resin. This report reviews the literature on the thermal and radiolytic degradation of ion exchange resins and provides calculations to estimate the fraction of volatile iodine as a function of time under postulated accident conditions

  15. 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

  16. Assessment of alternate ion exchange resins for improved antimony removal from the primary heat transport system

    Energy Technology Data Exchange (ETDEWEB)

    Burany, R.; Suryanarayan, S.; Husain, A. [Kinectrics, Inc., Toronto, ON (Canada)

    2015-07-01

    Radiation fields around the CANDU heat transport system are a major contributor to worker dose during inspection, maintenance and refurbishment activities. While Co-60 is typically the dominant contributor to radiation fields in CANDU reactors, Sb-124, an activation product of antimony, is also a significant contributor, accounting for 5-20% of the radiation fields. The goal of this research project was to investigate resins for improved removal of antimony under both oxidizing and reducing conditions.Several candidate resins were tested and short-listed through a sequence of iterative testing. The results of the laboratory testing have identified potential candidates for improved antimony removal. Further testing is required to ensure compatibility with existing station resin specifications. (author)

  17. Relationship between ion transport and the failure behavior of epoxy resin coatings

    International Nuclear Information System (INIS)

    Dong, Yuhua; Zhou, Qiong

    2014-01-01

    Highlights: •An epoxy resin-Q345 system with a sandwich structure was prepared. •Cl − ions permeated into epoxy resin coating prior to K + ions. •Free volume size and PAL increased when the coating was immersed into the solution. -- Abstract: An epoxy resin coating with a sandwich structure was prepared to investigate ion transport behavior in the coating. The macro- and micro- appearance of the coating immersed in 5 wt.% KCl solutions was observed by stereomicroscopy, scanning electron microscopy equipped with an energy dispersive spectrometer. The electrochemical property of the coating was characterized by electrochemical impedance spectroscopy, and change of free volume after immersion was characterized by positron annihilation lifetime spectroscopy. The results indicated that Cl − ions permeated into the coating prior to K + ions, the free volume size and positron annihilation lifetime of the coating increased during immersion

  18. Operation and maintenance of the T-3 cask system

    International Nuclear Information System (INIS)

    Hussey, M.W.; Berger, J.D.; Peterson, J.M.

    1983-01-01

    The T-3 cask system consists of three lead-shielded casks and the associated payload containers, internal fixturing, tiedowns, transportation trailers and handling devices. The three casks were designed to meet the requirements of Title 10 of the Code of Federal Regulations, Part 71. The Nuclear Regulatory Commission cask licensing activities for original design and for licensing revisions have required significant analytical support. Commercial transportation contractors can provide needed services including provisions of suitable equipment, compliances with security requirements, and safe movement of the shipment at a potential savings over DOE-owned transportation systems. Proper periodic inspection/maintenance activities supported by adequate decontamination facilities are a must in keeping the T-3 casks available for service

  19. Development of NUPAC 140B 100 ton rail/barge cask

    International Nuclear Information System (INIS)

    1990-04-01

    The 140-B Cask Ancillary Equipment includes all cask-related hardware necessary for a complete transportation package and for handling of the cask at shipping and receiving facilities. The transportation package equipment includes the cask tiedown system, the railcar and the sunshield/personnel barrier. The cask handling systems include both single and dual load path cask lifting fixtures, a cask uprighting system, an intermodal transfer system, and the cask drain and fill system. This document describes the individual systems in terms of their purpose, their function, and their mechanical features. Structural analyses are provided for the cask lifting and tiedown devices. The cask ancillary equipment will also include special tools and equipment such as seal surface protection device, special torque wrenches, leak test equipment, etc., for handling the cask at a reactor site. Although final design work remains to be completed, the ancillary equipment design information presented in this document ensures that the 140-B cask transportation package will meet or exceed all structural, functional, and operational requirements, within the specified gross vehicle weight limit. 18 figs

  20. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

    Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

    2003-01-01

    Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

  1. Experimental drop tests in the frame of the component testing of transport casks for radioactive materials; Durchfuehrung experimenteller Fallpruefungen im Rahmen der mechanischen Bauartpruefung von Transportbehaeltern fuer radioaktive Stoffe

    Energy Technology Data Exchange (ETDEWEB)

    Musolff, A.; Droste, B.; Quercetti, T.; Mueller, K. [BAM Bundesanstalt fuer Materialforschung und -pruefung, Berlin (Germany)

    2010-05-15

    The Bundesanstalt fuer Materialforschung und -pruefung (BAM) is the responsible German authority for component testing of packaging containers for radioactive materials in the frame of the licensing process. The tests are performed according to the national and international legal regulations for the transport of radioactive materials. This includes the evaluation of the mechanical and thermal behavior under defined testing conditions and the radionuclide release in case of severe accidents. The most important test concerning the accident resistance is the experimental drop tests from 9 m height on a firm base plate and from 1 m height on a steel mandrel. The tested sample was a 1:2 model cask CASTOR {sup registered} HAW/Tb2 of the original type CASTOR {sup registered} HAW28M. The tests were performed at the BAM test ground in Hortswalde. The documented test results will be part of the safety analysis in the frame of the licensing process for the transport and storage cask.

  2. 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.)

  3. 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)

  4. Loading 076 assemblies in two IV-04 transport casks for transport to the U.S. Savannah River Site (SC); Trasferimento di 72 elementi irraggiati MTR dalla piscina dell`impianto EUREX a due contenitori IU-04 per il trasporto al Savannah River Site-Department of Energy (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Gili, Michele [ENEA, Centro Ricerche Saluggia, Vercelli (Italy). Dipt. Energia

    1997-09-01

    The National Agency for New Technologies and the Environments has signed with the US Department of Energy a contract for the transfer of 150 irradiated MTR fuel assemblies stored in the EUREX plant pool at The National Agency for New Technologies and the Environments Research Centre of Saluggia. The first scheduled transport has been made in february 1997 and has involved the successful loading of 76 assemblies in two IU-04 (Pegase) transport casks. The loaded casks have been shipped to the U.S. Savannah River Site (SC).

  5. Study on leakage rates of high temperature water from wet-type transport casks for spent fuel. Pt. 2. Leakage rates from a scratch on O-ring surface and narrow wires adhering to O-ring surface

    International Nuclear Information System (INIS)

    Asano, R.; Aritomi, M.; Sudi, A.; Kohketsu, Y.

    1997-01-01

    A programme for enhancement of fuel burnup has been promoted in Japan as part of the sophisticated programme for light water reactors to reduce the fuel cost and the amount of spent fuel. As part of this fuel programme, a new wet-type transport cask has been developed to transport the high burnup fuels efficiently. The purpose of this work is to clarify the margin of safety in the evaluation of the release rate of radioactive materials from the wet-type transport cask into the environment and to establish a practical evaluation method for leakage rates on leak behaviour of high temperature water from the casks. In this paper, leakage rates of water under high pressures and at high temperatures are investigated from two kinds of leak path model. One is a disc with a scratch on the surface which simulates a defect on the seal surface of the O-ring flange and the other is narrow stainless steel wires installed on the O-ring surface which simulates hair adhering to the O-ring surface. From the results, an evaluation method for the leakage rate of water under high pressure and at high temperature from a non-circular leak path and multiple leak paths is proposed. (author)

  6. Development of safety-relevant components for the transport and handling of final storage casks for waste from decommissioning

    International Nuclear Information System (INIS)

    Bruening, D.; Geiser, H.; Kloeckner, F.; Rittscher, D.; Schlesinger, H.J.

    1992-10-01

    The aim of the study was the development, construction and testing of a transportation system that is able to transport cylindrical waste containers as well as containers from the deliverer to the 'KONRAD' final repository. A transport palette has been developed that can carry two cylindrical waste containers with type B requirement or classification II. An Open-All-Container for the transport of palettes and 'KONRAD' containers has been developed. A storage of cylindrical waste containers and containers in the final repository is possible with the newly developed transportation system. Safety specifications of the transportation system have been passed successfully. (orig.). 30 refs., 8 tabs., 74 figs [de

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

  9. 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)

  10. Developing cask designs in the USSR

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    To tackle the problem of transporting spent fuel from its VVER-1000s, the Soviet Union has developed two casks - the TK-10 and the TK-13 which are described here. Future developments of these designs may use a silicon-organics based material for the solid neutron shielding and no neutron absorbers in the fuel assembly basket. (author)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. Full-scale prototyping of the Hitachi dual-purpose metal cask and verification of its heat transfer characteristics

    International Nuclear Information System (INIS)

    Kumagai, N.; Ishida, N.; Ootsuka, M.; Kamoshida, M.; Hiranuma, T.; Doumori, S.; Hoshikawa, T.; Shimizu, M.; Kashiwakura, J.; Hayashi, M.

    2004-01-01

    Hitachi has been developing dual-purpose metal casks for transport and storage of spent nuclear fuels. The Hitachi cask, HDP69B can store 69 BWR fuel assemblies. The cask features are as follows. 1) The fuel basket is assembled mainly with plates of borated stainless steel. The plates are not welded, but cross-inserted into each other like the dividers in an egg carton. Since the borated stainless steel has relatively low heat conductivity, aluminum alloy plates are inserted along with some stainless steel plates to enhance heat removal ability. 2) Cured resin blocks are fitted into the inner shell of the cask for neutron shielding of the cask body. The resin blocks are surrounded by an aluminum casing which transfers heat of stored fuel from the inner shell to the outer shell of the cask. The block type shield structure eliminates the need for welding the heat transfer fins to the inner and outer shells. The weldless structures of the HDP69B lead to its enhanced manufacturability, but they complicate the heat transfer characteristics because there are gaps between such components as the aluminum casing and inner/outer shells. We carried out full-scale prototyping of the HDP69B and ran a heat transfer test using the prototype. The purposes of the heat transfer test were to check the heat removal ability of the HDP69B and to verify the safety analysis model for heat removal. Results of the heat transfer test and optimized analysis model for heat transfer characteristics of the HDP69B are the focus of this paper. The heat transfer test is summarized as follows. Sixty nine heaters simulating the shape and heat power of spent fuel assemblies were inserted into the fuel basket. After replacing the inner atmosphere with 0.1 MPa of helium, the heat transfer test was started. About 7 days were required to equilibrate the temperature distribution. The temperature at the center of the basket was 194 C. The results confirmed the HDP69B had sufficient heat removal ability. The

  18. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  19. 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)

  20. 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

  1. Decree no. 2001-1199 of the 10 december 2001 publishing the resolution MSC. 88 (71) notifying adoption of the international compilation of safety rules for the spent nuclear fuels, plutonium and high level radioactive wastes transport in casks on ships (compilation INF) (annexes), adopted at London the 27 may 1999; Decret no. 2001-1199 du 10 decembre 2001 portant publication de la resolution MSC.88 (71) portant adoption du recueil international de regles de securite pour le transport de combustible nucleaire irradie, de plutonium et de dechets hautement radioactifs en colis a bord de navires (recueil INF) (ensemble une annexe), adoptee a Londres le 27 mai 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This legislative text concerns the safety rules of spent nuclear fuels, plutonium and high level radioactive wastes transport, in casks on ships. Rules, fire prevention, temperature control of casks, electric supply, radioprotection, management and emergency plans are detailed. (A.L.B.)

  2. Cask handling method and apparatus

    International Nuclear Information System (INIS)

    Yoli, A.H.; Husain, I.

    1977-01-01

    The method of transferring radioactive material into and out of the cask comprises positioning a tank with an open end in a well. Then a cask having a passage for moving radioactive material into and out of the cask is placed in the tank through the opening in the tank. The tank opening is then sealed to the cask relative to the well without sealing the passage relative to the well to prevent water filled into the well from leaking into the tank. Then the well is filled with water above the seal, and radioactive material is then moved through the water in the well through the passage into the cask. The tank may be filled with demineralized water from a separate source to pressurize the space in the tank on the other side of the seal from the well to prevent water in the well from entering the tank. The water level in the well and in the tank is then lowered, the tank opening to the cask seal is removed, and a cover is attached to the cask passage to maintain the radioactive material and contaminated water in the cask. The apparatus which accomplishes the above method comprises a tank in a well for receiving a cask therein. A seal between the tank and the cask prevents water in the well from flowing into the tank about the cask and permits water in the well to flow through the cask opening into the cask. A first water supply means raises and lowers the water level in the well, and a second water supply means supplies clean demineralized water to the tank under pressure to prevent water in the well from leaking into the tank. The seal is annularly shaped and is attached to the top of the tank. The central portion of the annular seal is aligned with the cask opening and it has means to seal the annular seal to the cask

  3. Blemished image? Results of an Emnid study into the effects of recent spent fuel cask transports and the deregulation of the energy markets

    International Nuclear Information System (INIS)

    Hinzmann, U.

    1998-01-01

    Have the radioactivity leaks from spent nuclear fuel casks conveyed to waste management sites in Germany left a stain on the image of the electric power companies? The Bielefeld-based Emnid Institute, unit for energy market research, carried out an opinion poll among the population late in May and again in mid-June 1998, in oder to find out whether the leaking casks have blemished the image of the nuclear power industry. Another aspect inquired was the expectations of the population in connection with the deregulation of the electric power market. (orig./CB) [de

  4. Computerized simulation of the mechanical behavior of wood-filled shock absorbers of radioactive materials transport casks

    International Nuclear Information System (INIS)

    Neumann, Martin; Wille, Frank

    2011-01-01

    In Germany the mechanical component inspection of transport containers for radioactive materials is performed by BAM (Bundesanstalt fuer Materialforschung und -pruefung) under consideration of national and international standards and guidelines. Experimental and calculative (analytical and numerical) techniques combined with material and/or component testing are the basis of assessment concepts according the state of the art. The authors describe the experiences of BAM concerning assessment and description of the mechanical behavior of shock absorbing components, including modeling strategies, material models, drop tests and experiment-calculation comparison. Energy absorbing components are used to reduce the impact forces at the container in case of a transport accident. In Germany wood filled thin-walled constructions are used. The deformation behavior of the wood is a main part of the calculative simulation procedures in comparison with experimental tests.

  5. Spreader beam analysis for the CASTOR GSF cask

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

    The purpose of this report is to document the results of the 150% rated capacity load test performed by DynCorp Hoisting and Rigging on the CASTOR GSF special cask lifting beams. The two lifting beams were originally rated and tested at 20,000kg (44,000lb) by the cask manufacturer in Germany. The testing performed by DynCorp rated and tested the lifting beams to 30,000 kg (66,000 lb)+0%, -5%, for Hanford Site use. The CASTOR GSF cask, used to transport isotopic Heat Sources (canisters), must be lifted with its own designed lifting beam system (Figures 1, 2, and 3). As designed, the beam material is RSt 37-2 (equivalent to American Society for Testing and Materials[ASTM] A-570), the eye plate is St 52-2 (equivalent to ASTM A-516), and the lifting pin is St 50 (equivalent to ASTM A-515). The beam has two opposing 58 mm (2.3 in.) diameter by 120 mm(4.7 in.) length, high grade steel pins that engage the cask for lifting. The pins have a manual locking mechanism to prevent disengagement from the casks. The static, gross weight (loaded) of the cask 18,640 kg (41,000 lb) on the pins prevents movement of the pins during lifting. This is due to the frictional force of the cask on the pins when lifting begins

  6. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  7. Evaluation of impact limiter performance during end-on and slapdown drop tests of a one-third scale model storage/transport cask system

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Bronowski, D.R.; Uncapher, W.L.; Attaway, S.W.; Bateman, V.I.; Carne, T.G.; Gregory, D.L.; Huerta, M.

    1990-12-01

    This report describes drop testing of a one-third scale model shipping cask system. Two casks were designed and fabricated by Transnuclear, Inc., to ship spent fuel from the former Nuclear Fuel Services West Valley reprocessing facility in New York to the Idaho National Engineering Laboratory for a long-term spent fuel dry storage demonstration project. As part of the NRC's regulatory certification process, one-third scale model tests were performed to obtain experimental data on impact limiter performance during impact testing. The objectives of the testing program were to (1) obtain deceleration and displacement information for the cask and impact limiter system, (2) obtain dynamic force-displacement data for the impact limiters, (3) verify the integrity of the impact limiter retention system, and (4) examine the crush behavior of the limiters. Two 30-ft (9-m) drop tests were conducted on a mass model of the cask body and scaled balsa and redwood-filled impact limiters. This report describes the results of both tests in terms of measured decelerations, posttest deformation measurements, and the general structural response of the system. 3 refs., 32 figs

  8. SMART, Radiation Dose Rates on Cask Surface

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1989-01-01

    1 - Description of program or function: SMART calculates radiation dose rate at the center of each cask surface by using characteristic functions for radiation shielding ability and for radiation current back-scattered from cask wall and cask cavity of each cask, once cask-type is specified. 2 - Method of solution: Matrix Calculation

  9. Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

    International Nuclear Information System (INIS)

    Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-01-01

    A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper

  10. 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.)

  11. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  12. Development of cask body integrated with bottom plate

    International Nuclear Information System (INIS)

    Yoshida, Takuji; Sasaki, Tomoharu; Koyama, Yoichi; Kumagai, Yasuyuki; Watanabe, Yuichi; Takasa, Seiju

    2017-01-01

    The main parts of a metal cask for storage and transport of spent nuclear fuel consists of main body, neutron shield material and external cylinder. The forged main body has been manufactured as a cup shape by welding of 'forged body' and 'forged bottom plate' which are independently forged. JSW has developed the manufacturing technology of 'cask body integrated with bottom plate' which has no weld line with the goal of cost reduction, manufacturing period shortening and further reliability improvement. Manufacturing for the prototype of 'cask body integrated with bottom plate' has completed to verify mechanical properties and uniformity of the product which satisfy the specified values stipulated in JSME Code S FA1 2007 edition. Here, we report the manufacturing technology and obtained properties of 'cask body integrated with bottom plate'. (author)

  13. Impact response analysis of cask for spent fuel by dimensional analysis and mode superposition method

    International Nuclear Information System (INIS)

    Kim, Y. J.; Kim, W. T.; Lee, Y. S.

    2006-01-01

    Full text: Full text: Due to the potentiality of accidents, the transportation safety of radioactive material has become extremely important in these days. The most important means of accomplishing the safety in transportation for radioactive material is the integrity of cask. The cask for spent fuel consists of a cask body and two impact limiters generally. The impact limiters are attached at the upper and the lower of the cask body. The cask comprises general requirements and test requirements for normal transport conditions and hypothetical accident conditions in accordance with IAEA regulations. Among the test requirements for hypothetical accident conditions, the 9 m drop test of dropping the cask from 9 m height to unyielding surface to get maximum damage becomes very important requirement because it can affect the structural soundness of the cask. So far the impact response analysis for 9 m drop test has been obtained by finite element method with complex computational procedure. In this study, the empirical equations of the impact forces for 9 m drop test are formulated by dimensional analysis. And then using the empirical equations the characteristics of material used for impact limiters are analysed. Also the dynamic impact response of the cask body is analysed using the mode superposition method and the analysis method is proposed. The results are also validated by comparing with previous experimental results and finite element analysis results. The present method is simpler than finite element method and can be used to predict the impact response of the cask

  14. 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

  15. 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.

  16. Trunnions for spent fuel element shipping casks

    International Nuclear Information System (INIS)

    Cooke, B.

    1989-01-01

    Trunnions are used on spent fuel element shipping casks for one or more of a combination of lifting, tilting or securing to a transport vehicle. Within the nuclear transportation industry there are many different philosophies on trunnions, concerning the shape, manufacture, attachment, inspection, maintenance and repair. With the volume of international transport of spent fuel now taking place, it is recognized that problems are occurring with casks in international traffic due to the variance of the philosophies, national standards, and the lack of an international standard. It was agreed through the ISO that an international standard was required to harmonize. It was not possible to evolve an international standard. It was only possible to evolve an international guide. To evolve a standard would mean superseding any existing national standards which already cover particular aspects of trunnions i.e. deceleration forces imposed on trunnions used as tie down features. Therefore the document is a guide only and allows existing national standards to take precedence where they exist. The guide covers design, manufacture, maintenance, repair and quality assurance. The guide covers trunnions used on spent fuel casks transported by road, rail and sea. The guide details the considerations which should be taken account of by cask designers, i.e. stress intensity, design features, inspection and test methods etc. Manufacture, attachment and pre-service testing is also covered. The guide details user requirements which should also be taken account of, i.e. servicing frequency, content, maintenance and repair. The application of quality assurance is described separately although the principles are used throughout the guide

  17. 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

  18. Post test evaluation of a fire tested rail spent fuel cask

    International Nuclear Information System (INIS)

    Rack, H.J.; Yoshimura, H.R.

    1980-01-01

    Postmortem examination of a large rail-transported spent fuel shipping cask which had been exposed to a JP-4 fuel fire revealed the presence of two macrofissures in the outer cask shell. One, a part-through crack located within the seam weld fusion zone of the outer cask shell, is typical of hot cracks found in stainless steel weldments. The other, a through-crack, was apparently initiated during the formation of a copper-stainless steel dissimilar metal joint, with crack propagation through the cask outer shell having occurred during the fire-test. 8 figures

  19. Application of the ASME code in the design of the GA-4 and GA-9 casks

    International Nuclear Information System (INIS)

    Mings, W.J.; Koploy, M.A.

    1992-01-01

    General Atomics (GA) is developing two spent fuel shipping casks for transport by legal weight truck (LWT). The casks are designed to the loading, environmental conditions and safety requirements defined in Title 10 of the Code of Federal Regulations, Part 71 (10CFR71). To ensure that all components of the cask meet the 10CFR71 rules, GA established structural design criteria for each component based on NRC Regulatory Guides and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paper discusses the criteria used for different cask components, how they were applied and the conservatism and safety margins built into the criteria and assumption

  20. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    International Nuclear Information System (INIS)

    Diggs, J.M.

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables

  1. Investigation of the behaviour of impact limiting devices of transport casks for radioactive materials in the package approval and risk analysis; Untersuchung des Verhaltens stossdaempfender Bauteile von Transportbehaeltern fuer radioaktive Stoffe in Bauartpruefung und Risikoanalyse

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Martin

    2009-08-20

    Transport casks for radioactive materials with a Type-B package certificate have to ensure that even under severe accident scenarios the radioactive content remains safely enclosed, in an undercritical arrangement and that ionising radiation is sufficiently shielded. The impact limiter absorbs in an accident scenario the major part of the impact energy and reduces the maximum force applied on the cask body. Therefore the simulation of the behaviour of impact limiting devices of transport casks for nuclear material is of great interest for the design assessment in the package approval as well as for risk analysis in the field of transport of radioactive materials. The behaviour of the impact limiter is influenced by a number of parameters like impact limiter construction, material properties and loading conditions. Uncertainties exist for the application of simplified numerical tools for calculations of impact limiting devices. Uncertainities exist when applying simplified numerical tools. A model describing the compression of wood in axial direction of wood under large deformations for simulation with complex numerical procedures like dynamic Finite Element Methods has not been developed yet. Therefore this thesis concentrates on deriving a physical model for the behaviour of wood and analysing the applicability of different modeling techniques. A model describing the compression of wood in axial direction under large deformations was developed on the basis of an analysis of impact limiter of prototypes of casks for radioactive materials after a 9-m-drop-test and impact tests with wooden specimens. The model describes the softening, which wood under large deformation exhibits, as a function of the lateral strain constraint. The larger the lateral strain restriction, the more energy wood can absorb. The energy absorption capacity of impact limiter depends therefore on the ability of the outer steel sheet structure to prevent wood from evading from the main

  2. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  3. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  4. Research and development of spent fuel shipping casks and the criteria for seagoing vessel carrying casks

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Considering that the transportation of spent fuel will increase rapidly and extensively in the near future, Japanese Atomic Energy Committee enacted ''Technical Standard for Transportation of Radioactive Materials'' based on ''IAEA Regulation for the Safe Transport of Radioactive Materials 1973 Revised Edition''. Coping with the recommendation of AEC, Atomic Energy Bureau in Science and Technology Agency and other authorities concerned started to review the former ordinances for transportation of radioactive materials and to consolidate a unified system of relevant laws and standards. On the other hand, Atomic Energy Bureau has invested in research and development since ten years ago in order to obtain the data for design and licensing work of spent fuel shipping casks. In those studies some different scale models of a prototype of 80 t in weight have been used to make clear the scale effect at the drop, pucture and fire tests, which are one of the features of Japanese research and development. And also the immersion test in high pressure water up to about 500 bars is now carried out to investigate the integrity of cask body and sealing structure to prevent leakage of radioactive contents to the ambient when the cask falls into deep sea. In Japan, depending on the site conditions of nuclear plants, almost all transportations of unirradiated and spent fuels are done on the sea. Therefore, in order to secure the safety of transportation, the design criteria of the seagoing vessels for exclusive transportation of spent fuel shipping casks, namely full load shipping, has been enacted, which aims to make minimum the probability of sinking at collison, stranding and other unforeseen accidents at sea and also to restrain radiation exposure of the crew as low as possible

  5. Research on localization and alignment technology for transfer cask

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jingchuan, E-mail: jchwang@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China); Yang, Ming; Chen, Weidong [Department of Automation, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, Shanghai (China)

    2015-10-15

    Highlights: • A method for the alignment between TB and HCB based on localizability is proposed. • A localization method based on the localizability estimation is proposed to realize the cask's localization accurately and ensures the transfer cask's accurate docking in the front of the window of Tokmak Building. • The experimental results show that the proposed algorithm works well in the indoor simulation environment. This system will be test in EAST of China. - Abstract: According to the long length characteristics of transfer cask compared to the environment space between Tokmak Building (TB) and HCB (Hot Cell Building), this paper proposes an autonomous localization and alignment method for the internal components transportation and replacement. A localization method based on the localizability estimation is used to realize the cask's localization and navigation accurately. Once the cask arrives at the front of the TB window, the position and attitude measurement system is used to detect the relative alignment error between the seal door of pallet and the window of TB real-time. The alignment between seal door and TB window could be realized based on this offset. The simulation experiment based on the real model is designed according to the real TB situation. The experiment results show that the proposed localization and alignment method can be used for transfer cask.

  6. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C... Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Title: Topical Safety Analysis Report for the NAC Storage/Transport Cask for Use at an Independent Spent...

  7. 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)

  8. 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...

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

    International Nuclear Information System (INIS)

    Andress, D.; McLeod, N.B.; Rahimi, M.; Joy, D.S.; Peterson, R.W.

    1991-01-01

    The DOE has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design - the GA-4 and GA-9 truck casks and the BR-100 rail cask. The GA-4 cask is designed for PWR fuel only; the GA-9 cask is a longer cask with less shielding designed for BWR fuel only; and the BR-100 cask is designed to accommodate both PWR and BWR fuels. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elements as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. Because of the button and spring interference, the basket openings in these casks will not accommodate assemblies in the BWR/2,3 and BWR/4-6 fuel classes with the fuel channels in place

  10. Impact analysis of shipping casks

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1989-01-01

    This paper describes how simpler two- and three-dimensional models can be used to provide an intermediate level of detail between full three dimensional finite element calculations and hand calculation. Three free drop scenarios are analyzed to assess the integrity of the cask when subjected to large bending and axial stresses. These three drop scenarios are: a thirty foot axial drop on either end, a thirty foot oblique angel drop with the cask having several different orientations from the vertical with impact on the top end cask corner, and a thirty foot side drop with simultaneous impact on one of the lifting trunnions and the bottom end. Prevention of damage hinges on the strength of the various components that comprises the cask. The predicted levels of deformation and stresses in the cask are used to assess the potential damage level

  11. Calculation of radiation dose rates from a spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask

  12. Scale-4 shipping cask shielding applications

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Parks, C.V.

    1991-01-01

    This paper reports the application of the SCALE-4 shielding sequences SAS1 and SAS4 to the problem set distributed by the Organization for Economic Cooperation and Development (OECD) Working Group on Shielding Assessment of Transportation Packages. In many cases, additional comparison are made with MCNP and QADS solutions to provide a complete cross-check of methods, cross sections, geometry, etc. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. Bituminous solidification, disposal, transport and burial of spent ion-exchange resins. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Mozes, G.; Kristof, M.

    1983-07-01

    The project dealing with the incorporation of spent ion-exchange resins into bitumen was performed within the Agency coordinated research programme on treatment of spent ion-exchange resins. Physical and chemical properties of commercial ion-exchange resins, bitumens and bituminized resins were studied. It was shown that bitumen with low oil content and with a softening point of 60-70 deg. C are applicable for the incorporation of resins. The final waste form is allowed to contain maximum 50% resin. The comprehensive study of the biological resistance of B-30 bitumen was performed. That showed that any bacteriological attack can be regarded as generally insignificant. A continuously operating technology was realized on a semi-plant scale. The best operating conditions of this technology were determined. On the basis of the experience gained from the experiments a design of the bituminization plant of 50m 3 dry resin/year treatment capacity was proposed

  19. Thermal Evaluation of a KRI-BGM Shipping Cask

    International Nuclear Information System (INIS)

    Bang, K. S.; Lee, J. C.; Seo, K. S.

    2007-01-01

    Radioactive isotopes are used extensively in the fields of industry, medical treatment, food and agriculture. Use of radioactive isotopes is expected to increase continuously with the growth of each field. In order to safely transport radioactive isotopes from the place of manufacture to the place of use, a shipping package is required. Therefore KAERI is developing the KRI-BGM shipping cask to transport the Ir-192 bulk radioactive material, which is produced at the HANARO research reactor. The shipping package should satisfy the requirements which are prescribed in the Korea MOST Act 2001-23, IAEA Safety Standard Series No. TS-R-1, US 10 CFR Part 71 and the US 49 CFR Part 173. These regulatory classify the KRI-BGM shipping cask as a Type B package, and their regulatory guidelines state that the Type B package for transporting radioactive materials should be able to withstand a period of 30 minutes under a thermal condition of 800 .deg. C. However, the polyurethane, which is to be used as the filling within the cavity of the KRIBGM shipping cask, has a very weak characteristic in a high temperature. Therefore it is difficult for the depleted uranium(hereafter DU), which is used as shielding material, to be protected under a thermal condition of 800 .deg. C. Accordingly, the KRI-BGM shipping cask, which applied non-combustible polyurethane and fireproof materials as the filling, was fabricated. The thermal tests by using prototype cask have been performed to estimate the thermal integrity of the KRI-BGM shipping cask under a thermal condition of 800 .deg. C

  20. Prototypical fabrication of PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Kwack, Eun Ho; Kim, Byung Ku; Kang, Hee Yung; Lee, Chung Young; Jeon, Kyeong Lak; Lee, Bum Soo

    1985-02-01

    This report describes about the safety analysis for the spent fuel shipping cask, which is used to transfer a single fuel assembly discharged from PWR in operation in Korea. The contents cover the methods and the results of structural, thermal, thermo-hydraulic, radiation shield and criticality detail analysis. The safety evaluation has been made under the normal transportation and hypothetical accident conditions such as 30ft free drop, puncture, fire, immersion, penetration, corner drop, etc,. Some corrections in design are made, and a brief information for fabrication and transportation are obtained by the use of a 1/6 scale model. The design is based on one year cooling time of the spent fuel with 40,000 MWT/MTU maximum burnup, which gives 7.2KW decay heat and 1.6x10 6 ci/hr radiation intensity. The cask is composed of main body with the double closures, impact limiter and fuel basket. The inner shell, inner closure and valves constitute the pressure boundary of the containment. The inner, intermediate and outer shells, upper and lower forgings are made of stainless steel which compose the main body with lead for gamma shield and 50% ethylene glycol for neutron shield. The impact limiters are made of balsa wood on both end sides of the cask to protect the cask from a sudden shocks in accident during the transportation. The analysis results show that the cask is proved to retain its structural integrity within allowable stress and to be safe under the normal and hypothetical accident conditions, and the maximum dose rates of radiation at 2m distance from the surface of the cask are less than the required values. The weight will be 23.2tons in dry and 27.8 tons in wet with fuel loaded. All the design data, calculated results for the structural integrity, shield and thermal analysis are shown in this report with the basic drawings. (Author)

  1. 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

  2. Impact of an exploding LPG rail tank car onto a CASTOR spent fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Droste, B.; Probst, U.; Heller, W

    1999-07-01

    On 27 April 1999 a fire test was performed with a 45 m{sup 3} rail tank car partially filled with 10 m3 pressurised liquid propane. A CASTOR THTR/AVR spent fuel transport cask was positioned beside the propane tank as to suffer maximum damage from any explosion. About 17 min after fire ignition the propane tank ruptured. This resulted in a BLEVE with an expanding fireball, heat radiation, explosion overpressure, and tank fragments projected towards the cask. This imposed severe mechanical and thermal impacts directly onto the CASTOR cask, moving it 17 m from its original position. This involved rotation of the cask with the lid end travelling 10 m before it crashed into the ground. Post-test investigations of the CASTOR cask demonstrated that no loss of leaktightness or containment and shielding integrity occurred. (author)

  3. A revision of the cask designers guide for the '90s

    International Nuclear Information System (INIS)

    Shappert, L.B.; Green, V.M.

    1993-01-01

    DOE has requested that ORNL initiate a revision to NSIC-68, A Guide for the Design, Fabrication, and Operation of Shipping Casks for Nuclear Applications, commonly called the Cask Designers Guide. This revision, called the Cask Handbook, has two goals: (1) to improve the quality of SARPs that are submitted to DOE, and (2) to provide up-to-date information on the design of spent fuel shipping casks, including information on fabrication, quality assurance, SARP preparation, certification, use, maintenance, and other general topics. The revision provides guidance that will help engineers through the cask licensing process, in part, by providing as much regulator-approved data and 'lessons-learned' information as possible. The effort is sponsored by DOE-Environmental, Safety and Health (EH), guided by Transportation Technology staff members at ORNL, and the information is being generated by experts in the various technical fields. (J.P.N.)

  4. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  5. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N.

    2004-01-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO 2 , Al 2 O 3 , Gd 2 O 3 , etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO 2 ) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al 2 O 3 ) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO 2 for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate casks with variable cermet compositions

  6. Cask size and weight reduction through the use of depleted uranium dioxide-concrete material

    International Nuclear Information System (INIS)

    Lobach, S.Yu.; Haire, J.M.

    2007-01-01

    Newly developed depleted uranium (DU) composite materials enable fabrication of spent nuclear fuel (SNF) transport and storage casks that are smaller and lighter in weight than casks made with conventional materials. One such material is DU dioxide (DUO2)-concrete, so-called DUCRETE TM . This paper examines the radiation shielding efficiency of DUCRETE as compared with that of a conventional concrete cask that holds 32 pressurized-water-reactor SNF assemblies. In this analysis, conventional concrete shielding material is replaced with DUCRETE. The thickness of the DUCRETE shielding is adjusted to give the same radiation surface dose, 200 mrem/h (2 mSv/hr), as the conventional concrete cask. It was found that the concrete shielding thickness decreased from 71 to 20 cm and that the cask radial cross-section shielding area was reduced approx 50 %. The weight was reduced approx 21 %, from 154 to approx 127 tons. Should one choose to add an extra outer ring of SNF assemblies, the number of such assemblies would increase from 32 to 52. In this case, the outside cask diameter would still decrease, from 169 to 137 cm. However, the weight would increase somewhat from 156 to 177 tons. Neutron cask surface dose is only approx 10 % of the gamma dose. These reduced sizes and weights will significantly influence the design of next-generation SNF casks

  7. Safety analysis report for packaging: neutron shipping cask, model 0.5T

    International Nuclear Information System (INIS)

    Peterson, R.T.

    1976-01-01

    The Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The shipping cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual

  8. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    International Nuclear Information System (INIS)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il'kaev, R.I.; Shapovalov, V.I.

    2004-01-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks

  9. Estimation of terrorist attack resistibility of dual-purpose cask TP-117 with DU (depleted uranium) gamma shield

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, O.G.; Matveev, V.Z.; Morenko, A.I.; Il' kaev, R.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)

    2004-07-01

    Report is devoted to numerical research of dual-purpose unified cask (used for SFA transportation and storage) resistance to terrorist attacks. High resistance of dual-purpose unified cask has been achieved due to the unique design-technological solutions and implementation of depleted uranium in cask construction. In suggested variant of construction depleted uranium fulfils functions of shielding and constructional material. It is used both in metallic and cermet form (basing on steel and depleted uranium dioxide). Implementation of depleted uranium in cask construction allows maximal load in existing overall dimensions of the cask. At the same time: 1) all safety requirements (IAEA) are met, 2) dual-purpose cask with SFA has high resistance to terrorist attacks.

  10. Structural challenges in the development of a truck shipping cask for the OCRWM Cask Systems Development Program

    International Nuclear Information System (INIS)

    Mello, R.M.; Severson, W.J.; Nair, B.R.

    1990-01-01

    The development of a spent fuel transportation cask design based on a structural material without licensing precedent presents many challenges. The US Nuclear Regulatory Commission (NRC) requires that any new material be qualified to meet the design and fabrication requirements of the ASME Boiler ampersand Pressure Vessel Code, Section III, Class 1. This paper discusses the strategy that is being implemented towards obtaining Code acceptance of a titanium alloy (3A1-2.5V). This alloy has been chosen as the principal structural material for a Legal Weight Truck cask being developed by Westinghouse for the US Department of Energy. The analysis approach used on some of the principal cask components is also presented. 5 refs., 8 figs., 3 tabs

  11. Structural challenges in the development of a truck shipping cask for the OCRWM cask systems development program

    International Nuclear Information System (INIS)

    Mello, R.M.; Severson, W.J.; Nair, B.R.

    1990-01-01

    The development of a spent fuel transportation cask design based on a structural material without licensing precedent presents many challenges. The U.S. Nuclear Regulatory Commission (NRC) requires that any new material be qualified to meet the design and fabrication requirements of the ASME Boiler and Pressure Vessel Code, Section III, Class 1. This paper discusses the strategy that is being implemented towards obtaining Code Acceptance of a titanium alloy (3A1-2.5V). This alloy has been chosen as the principal structural material for a Legal Weight Truck cask being developed by Westinghouse for the U.S. Department of Energy. The analysis approach used on some of the principal cask components is also presented

  12. 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...

  13. 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...

  14. Development of Neutron Shielding Material for Cask and Accelerator

    International Nuclear Information System (INIS)

    Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

    2008-01-01

    The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

  15. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. (Audin (Lindsay), Ossining, NY (United States))

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  16. Nuclear cask testing films misleading and misused

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L. [Audin (Lindsay), Ossining, NY (United States)

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.

  17. Nuclear cask testing films misleading and misused

    International Nuclear Information System (INIS)

    Audin, L.

    1991-10-01

    In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ''proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests

  18. Postmortem metallurgical examination of a fire-exposed spent fuel shipping cask

    International Nuclear Information System (INIS)

    Rack, H.J.; Yoshimura, H.R.

    1980-04-01

    A potmortem examination of a large fire-exposed rail-transported spent fuel shipping container has revealed the presence of two macrofissures in the outer cask shell. The first, a part-thru crack located within the seam weld fusion zone of the outer cask shell, was typical of hot cracks that may be found in stainless steel weldments. The second, located within the stainless steel base metal, apparently originated at microcracks formed during the welding of a copper-stainless steel dissimilar metal joint. The latter microcrack then propagated during the fire-test, ultimately penetrating the outer shall of the cask. 18 figures, 2 tables

  19. 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

  20. In vitro comparison rate of dental root canal transportation using two single file systems on the simulated resin blocks

    Directory of Open Access Journals (Sweden)

    Mohammad Javad Etesami

    2016-07-01

    Full Text Available Background and Aims: Cleaning and shaping is one of the most important stages in endodontic treatment. Single-file systems save time and reduce the risk of transmission of pathogens. This in vitro study was aimed to compare the rate of canal transportation after the preparation of the stimulated resin root canal with two single-file systems, namely Waveone and Reciproc. Materials and Methods: Thirty stimulated resin root canal blocks with size 8/0. 02 K file were randomly divided into two study groups. The preparation in Group A and Group B was performed using Reciproc and Waveone files, respectively. Pre and post- preparation photographs were taken and the images were superimposed to evaluate the inner and outer wall’s curvature tendency at three points (apical, middle and coronal using AutoCad pragram. Data were analyzed using T-test. Results: Based on the results, the degree of transportation in the inner and outer walls of the canal was less at the level of 3 millimeters (P0.05. Conclusion: Waveone showed better performance in the middle third of canal and this system maybe recommended.

  1. Nuclear transport

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    During january and february 2003, a unique event concerning nuclear transport was reported and rated 1 on the INES scale. This event concerns the absence of a maintenance operation on a shipping cask. This shipping cask was used for several years for nuclear transport inside La-hague site before being re-assigned to transport on public thoroughfare. The re-assignment of the cask should have been preceded and conditioned by a maintenance operation whose purpose is to check the efficiency of its radiation shield. During this period 2 on-site inspections concerning the transport of nuclear materials were performed. (A.C.)

  2. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

  3. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  4. Spent fuel shipping cask sealing concepts

    International Nuclear Information System (INIS)

    Sonnier, C.S.

    1989-05-01

    In late 1985, the International Atomic Energy Agency (IAEA) requested the US Program for Technical Assistance to IAEA Safeguards (POTAS) to provide a study which examined sealing concepts for application to spent fuel shipping casks. This request was approved, and assigned to Sandia National Laboratories (Sandia). In the course of this study, discussions were held with personnel in the International Safeguards Community who were familiar with the shipping casks used in their States. A number of shipping casks were examined, and discussions were held with two shipping cask manufacturers in the US. As a result of these efforts, it was concluded that the shipping casks provided an extremely good containment, and that many of the existing casks can be effectively sealed by applying the seal to the cask closure bolts/nuts

  5. The metal and concrete cask for SNF and its radiation protection quality control

    International Nuclear Information System (INIS)

    Shchigolev, N.D.; Golubev, O.M.

    2005-01-01

    The transportation and packing module on the basis of metal and concrete cask for the long-term storage and shipment of the spent nuclear fuel is developed in compliance with the requirements of the national standards and IAEA recommendations. Such wares designed for the NPP and submarine reactors fuel may be remade also for the research ones. A procedure and remote device for the radiation protection control of this cask equally its integrity checks after dynamic testing also is described. (author)

  6. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed

  7. Gamma ray benchmark on the spent fuel shipping cask TN 12

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Cladel, C.; Ermont, G.; Nimal, J.C.

    1983-05-01

    The purpose of this benchmark is to compare measurements and calculation of gamma-ray dose rates around a shipping cask loaded with 12 spent fuel elements of FESSENHEIM PWR type. The benchmark provides a means to verify gamma-ray sources and gamma-ray transport calculation methods in shipping cask configurations. The comparison between measurements and calculations shows a good agreement except near the fuel element top where the discrepancy reaches a factor 2

  8. 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)

  9. Evaluation of RAPID for a UNF cask benchmark problem

    Science.gov (United States)

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

    2017-09-01

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

  10. Evaluation of RAPID for a UNF cask benchmark problem

    Directory of Open Access Journals (Sweden)

    Mascolino Valerio

    2017-01-01

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

  11. Fire resistant nuclear fuel cask

    International Nuclear Information System (INIS)

    Heckman, R.C.; Moss, M.

    1979-01-01

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

  12. Two microcephaly-associated novel missense mutations in CASK specifically disrupt the CASK-neurexin interaction.

    Science.gov (United States)

    LaConte, Leslie E W; Chavan, Vrushali; Elias, Abdallah F; Hudson, Cynthia; Schwanke, Corbin; Styren, Katie; Shoof, Jonathan; Kok, Fernando; Srivastava, Sarika; Mukherjee, Konark

    2018-03-01

    Deletion and truncation mutations in the X-linked gene CASK are associated with severe intellectual disability (ID), microcephaly and pontine and cerebellar hypoplasia in girls (MICPCH). The molecular origin of CASK-linked MICPCH is presumed to be due to disruption of the CASK-Tbr-1 interaction. This hypothesis, however, has not been directly tested. Missense variants in CASK are typically asymptomatic in girls. We report three severely affected girls with heterozygous CASK missense mutations (M519T (2), G659D (1)) who exhibit ID, microcephaly, and hindbrain hypoplasia. The mutation M519T results in the replacement of an evolutionarily invariant methionine located in the PDZ signaling domain known to be critical for the CASK-neurexin interaction. CASK M519T is incapable of binding to neurexin, suggesting a critically important role for the CASK-neurexin interaction. The mutation G659D is in the SH3 (Src homology 3) domain of CASK, replacing a semi-conserved glycine with aspartate. We demonstrate that the CASK G659D mutation affects the CASK protein in two independent ways: (1) it increases the protein's propensity to aggregate; and (2) it disrupts the interface between CASK's PDZ (PSD95, Dlg, ZO-1) and SH3 domains, inhibiting the CASK-neurexin interaction despite residing outside of the domain deemed critical for neurexin interaction. Since heterozygosity of other aggregation-inducing mutations (e.g., CASK W919R ) does not produce MICPCH, we suggest that the G659D mutation produces microcephaly by disrupting the CASK-neurexin interaction. Our results suggest that disruption of the CASK-neurexin interaction, not the CASK-Tbr-1 interaction, produces microcephaly and cerebellar hypoplasia. These findings underscore the importance of functional validation for variant classification.

  13. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-04-13

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube.

  14. Simulation of Multi Canister Overpack (MCO) Handling Machine Impact with Cask and MCO During Insertion into the Transfer Pit (FDT-137)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    The K-Basin Cask and Transportation System will be used for safely packaging and transporting approximately 2,100 metric tons of unprocessed, spent nuclear fuel from the 105 K East and K West Basins to the 200 E Area Canister Storage Building (CSB). Portions of the system will also be used for drying the spent fuel under cold vacuum conditions prior to placement in interim storage. The spent nuclear fuel is currently stored underwater in the two K-Basins. The K-Basins loadout pit is the area selected for loading spent nuclear fuel into the Multi-Canister Overpack (MCO) which in turn is located within the transportation cask. This Cask/MCO unit is secured.in the pit with a pail load out structure whose primary function is lo suspend and support the Cask/MCO unit at.the desired elevations and to protect the unit from the contaminated K-Basin water. The fuel elements will be placed in special baskets and stacked in the MCO that have been previously placed in the cask. The casks will be removed from the K Basin load out areas and taken to the cold vacuum drying station. Then the cask will be prepared for transportation to the CSB. The shipments will occur exclusively on the Hanford Site between K-Basins and the CSB. Travel will be by road with one cask per trailer. At the CSB receiving area the cask will be removed from the trailer. A gantry crane will then move the cask over to the transfer pit and load the cask into the transfer pit. From the transfer pit the MCO will be removed from the cask by the MCO Handling Machine (MHM). The MHM will move the MCO from the transfer pit to a canister storage tube in the CSB. MCOs will be piled two high in each canister Storage tube

  15. 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...

  16. 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

  17. 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

  18. Safety Analysis Report: Packages, Pu oxide and Am oxide shipping cask: Packaging of fissile and other radioactive materials: Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1984-12-01

    The PuO 2 cask or 5320-3 cask is designed for shipment of americium or plutonium by surface transportation modes. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, dated 12/21/76, which invokes Title 10 Code of Federal Regulations, Part 71 (10 CFR 71) ''Packaging of Radioactive Materials for Transport,'' and Title 49 CFR Parts 171.179 ''Hazardous Materials Regulations.'' (US DOE Order 4580.1A, Chapter III, superseded manual chapter 0529 effective May 1981, but it retained the same 10 CFR 71 and 49 CFR 171-179 references

  19. Design lead shielded casks for shipment and spent fuel from power reactors to reprocessing plant at Tarapur

    International Nuclear Information System (INIS)

    Seetharamaiah, P.

    1975-01-01

    Spent fuels from the Tarapur and Rajasthan Atomic Power Stations (TAPS and RAPS) are shipped to Fuel Reprocessing Plant at Tarapur in heavily lead shielded casks weighing about 65 tonnes as they are highly radioactive. The design of the casks has to meet stringemt requirements of safety and the integrity should be ensured to contain activity under credible accidents during handling and transportation. The paper presents the design of two casks for TAPS and RAPS spent fuel transportation particularly with reference to stress analysis considerations. The analysis also includes the handling gadgets and tie down attachments on the rail wagon and road trailer. (author)

  20. Heat transfer investigations within dry spent fuel casks

    International Nuclear Information System (INIS)

    Nitsche, F.

    1986-07-01

    For studying the heat transfer processes and predicting the maximum spent fuel element surface temperature in a spent fuel assembly (SFA) transported in a dry cask, model experiments have been performed with a gas-filled model cask containing a simplified electrically heated model of a WWER-type SFA with 90 fuel elements. The temperature distribution of the SFA model is measured for different heat rates under vacuum in the model cask, and under normal pressure and overpressure (0.1 ... 0.7 MPa) for several cooling gases (air, argon, helium) in order to separately investigate heat transfer processes by radiation and convection/conduction. The measuring results were compared with the calculations. Computer programmes as well as simplified calculation methods for temperature prediction were developed and checked. The results obtained are also useful for thermal analyses in the field of the dry storage of SFAs in a cask or can. Specifically it was found that: The heat removal from the SFA can be considerably improved by increasing the internal cask pressure or by using helium as coolant. The radiant heat exchange in the SFA model can be calculated with sufficient accuracy by means of a computer programme developed in 1978 or by means of a simplified analytical representation shown in the final report. Both methods are directly applicable to the original SFA and useful in order to approximately calculate the maximum SFE surface temperature under normal pressure, if the fraction of heat transferred by radiation is allowed for. For the calculation of the total heat transfer a computer programme was developed and verified, which completely permits the temperature prediction of the SFA model in dependence on heat rate, type of gaseous coolant and coolant pressure. This computer programme can be directly applied to the original SFA for the calculation of the maximum SFE surface temperature

  1. Criticality safety of spent fuel casks considering water inleakage

    International Nuclear Information System (INIS)

    Osgood, N.L.; Withee, C.J.; Easton, E.P.

    2004-01-01

    A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

  2. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-08-01

    Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs

  3. Estimate of the crud contribution to shipping cask containment requirements

    International Nuclear Information System (INIS)

    Sandoval, R.P.; Einziger, R.E.; Jordan, H.; Malinauskas, A.P.; Mings, W.J.

    1992-01-01

    This paper reports that a methodology is developed to relate U.S. Code of Federal Regulations, Title 10, Part 71 (10CFR71) containment requirements to leak rates for the special case in which the only radioactive species having a potential for escape form the cask is that associated with debris (crud) contained on the fuel assemblies being transported. The methodology accounts for the characteristics of the crud and for attenuation of the gas-borne crud particulates once they become suspended within the cask. Calculations are performed for typical spent-fuel transport cask geometries and the normal and accident conditions prescribed in 10CFR71. The most current published data are used for crud composition and structure, specific activity, spallation mechanics and fractions, and crud particle size. The containment criteria leak rates are calculated assuming 5-yr-old spent fuel. In each accident case, the containment leak rate criteria are well in excess of 10 cm 3 /s. Under normal conditions of transport, the regulatory containment requirements are met by leak rates ranging from 1.5 x 10 -3 cm 3 /s to 1.5 x 10 -4 cm 3 /s for the transport of boiling water reactor fuel assemblies and form 1.8 x 10 -2 cm 3 /s to 1.3 x 10 -3 cm 3 /s for pressurized water reactor fuel assemblies. The calculated leak rates are most sensitive to the cask design, type of fuel, and particle size distribution. Conservatism of the limiting leak rates is discussed

  4. 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.

  5. High-burnup/low-cooling-time fuel carrying capacity of the GA-4 and GA-9 spent fuel shipping casks

    International Nuclear Information System (INIS)

    Boshoven, J.K.; Hopf, J.E.

    1994-01-01

    In response to utilities' projected needs to ship higher burnup spent fuel, General Atomics (GA) has performed shielding and thermal analysis for the GA-4 and GA-9 legal weight shipping casks to determine the minimum cooling times for various burnup levels for fully loaded GA-4 and GA-9 casks and reduced payloads for the casks. Tables are provided in the paper which show the minimum cooling time for a given burnup and payload for each of the casks. The analyses show that the GA-4 and GA-9 casks can carry at least as many high-burnup and/or short-cooling-time spent fuel assemblies as present day shipping casks. In addition, the GA casks are able to carry at least twice as many assemblies as the present day shipping casks if the spent fuel burnup levels and/or cooling times are open-quotes coolerclose quotes or open-quotes as coolclose quotes as their design basis fuels. The increased shipping capacity for these more common open-quotes coolerclose quotes assemblies allows fewer shipments and therefore increases the efficiency and lowers predicted risks of the transport system

  6. 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

  7. 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...

  8. 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

  9. Comparison of apical transportation and change of working length in K3, NRT AND PROFILE rotary instruments using transparent resin block

    Directory of Open Access Journals (Sweden)

    Min-Jung Yoon

    2011-01-01

    Full Text Available Objectives The purpose of this study is to compare the apical transportation and working length change in curved root canals created in resin blocks, using 3 geometrically different types of Ni-Ti files, K3, NRT, and Profile. Materials and Methods The curvature of 30 resin blocks was measured by Schneider technique and each groups of Ni-Ti files were allocated with 10 resin blocks at random. The canals were shaped with Ni-Ti files by Crown-down technique. It was analyzed by Double radiograph superimposition method (Backman CA 1992, and for the accuracy and consistency, specially designed jig, digital X-ray, and CAD/CAM software for measurement of apical transportation were used. The amount of apical transportation was measured at 0, 1, 3, 5 mm from 'apical foramen - 0.5 mm' area, and the alteration of the working length before and after canal shaping was also measured. For statistics, Kruskal-Wallis One Way Analysis was used. Results There was no significant difference between the groups in the amount of working length change and apical transportation at 0, 1, and 3 mm area (p = 0.027, however, the amount of apical transportation at 5 mm area showed significant difference between K3 and Profile system (p = 0.924. Conclusions As a result of this study, the 3 geometrically different Ni-Ti files showed no significant difference in apical transportation and working length change and maintained the original root canal shape.

  10. 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

  11. 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.

  12. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  13. 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.

  14. 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)

  15. Safety analysis report for packaging: the ORNL tungsten-shielded cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Levine, D.L.; Just, R.A.

    1977-10-01

    The ORNL tungsten-shielded cask was designed and fabricated at the Oak Ridge National Laboratory (ORNL) for the transport of Type B and large quantity nonfissile isotopes. The container was evaluated analytically to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported. Computational methods were employed in a determination of the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and to the standards for the hypothetical accident conditions. The results of these evaluations demonstrate that the container is in compliance with the applicable regulations

  16. Multi-purpose canister storage unit and transfer cask thermal analysis

    International Nuclear Information System (INIS)

    Montgomery, R.A.; Niemer, K.A.; Lindner, C.N.

    1997-01-01

    Spent Nuclear Fuel (SNF) generated at commercial nuclear power plants throughout the US is a concern because of continued delays in obtaining a safe, permanent disposal facility. Most utilities maintain their SNF in wet storage pools; however, after decades of use, many pools are filled to capacity. Unfortunately, DOE's proposed final repository at Yucca Mountain is at least 10 years from completion, and commercial power utilities have few options for SNF storage in the interim. The Multi-Purpose Canister (MPC) system, sponsored by DOE's Office of Civilian Radioactive Waste Management, is a viable solution to the interim storage problem. The system is designed for interim dry storage, transport, and ultimate disposal of commercial SNF. The MPC system consists of four separate components: an MPC, Transfer Cask, Storage Unit, and Transport Cask. The SNF assemblies are loaded and sealed inside the helium-filled steel MPC. Once sealed, the MPC is not reopened, eliminating the need to re-handle the individual spent fuel assemblies. The MPC is transferred, using the MPC Transfer Cask, into a cylindrical, reinforced-concrete Storage Unit for on-site dry storage. The MPC may be removed from the Storage Unit at any time and transferred into the MPC Transport Cask for transport to the final repository. This paper discusses the analytical approach used to evaluate the heat transfer characteristics of an MPC containing SNF assemblies in the MPC Transfer Cask and Storage Unit

  17. 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

  18. 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

  19. Modal analysis of spent fuel cask for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Azimfar, S. A.; Kazemi, A.

    2011-01-01

    The Spent Fuel Assemblies of WWER-1000 reactors are planned to be transported by special containers which are supposed to be designed in a manner to stand against vibrations and impacts in order to protect the spent fuel from any possible damage. The vibration opposition of these containers shall be far beyond the critical resonance, because the resonances about the natural frequency of the structure will cause the enhancement of its oscillation range and may end with its disintegration. Determination of the amounts of natural frequencies and their mode shape can be achieved by vibration analyzing methods. The amount of the natural frequency of any structure crucially depends on its shape, material and lean points as well as the amount of the loads and the type of these loads. Due to the fact that the Spent Fuel Casks used for transportation in nuclear power plants in Russian Federation are TK-13 type and the pieces of information released are negligible, the scientists in Russia are working on the design and analysis of a new type made up of composite Material. In the presented paper the cask of spent fuel of TK-13 is modeled by ANSYS at 10.0 and ten natural frequency modes have been calculated, followed by the comparison of this result with the composite cask.

  20. Legal weight truck cask model impact limiter response

    International Nuclear Information System (INIS)

    Meinert, N.M.; Shappert, L.B.

    1989-01-01

    Dynamic and quasi-static quarter-scale model testing was performed to supplement the analytical case presented in the Nuclear Assurance Corporation Legal Weight Truck (NAC LWT) cask transport licensing application. Four successive drop tests from 9.0 meters (30 feet) onto an unyielding surface and one 1.0-meter (40-inch) drop onto a scale mild steel pin 3.8 centimeters (1.5 inches) in diameter, corroborated the impact limiter design and structural analyses presented in the licensing application. Quantitative measurements, made during drop testing, support the impact limiter analyses. High-speed photography of the tests confirm that only a small amount of energy is elastically stored in the aluminum honeycomb and that oblique drop slapdown is not significant. The qualitative conclusion is that the limiter protected LWT cask will not sustain permanent structural damage and containment will be maintained, subsequent to a hypothetical accident, as shown by structural analyses

  1. 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.)

  2. Incentives for the allowance of burnup credit in the design of spent nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-01-01

    An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in considerable public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities

  3. 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

  4. Analysis of radiation measurement data of the BUSS cask

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tang, J.S.

    1995-01-01

    The Beneficial Uses Shipping System (BUSS) is a Type-B packaging developed for shipping nonfissile, special-form radioactive materials to facilities such as sewage, food, and medical-product irradiators. The primary purpose of the BUSS cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for its certified special-form contents under both normal transport and hypothetical accident conditions. A BUSS cask that contained 16 CsCl capsules (2.723 x 10 4 TBq total activity) was recently subjected to radiation survey measurements at a Westinghouse Hanford facility, which provided data that could be used to validate computer codes. Two shielding analysis codes, MICROSHIELD (User's Manual 1988) and SAS4 (Tan 1993), that are used at Argonne National Laboratory to evaluate the safety of packaging of radioactive materials during transportation, have been selected for analysis of radiation data obtained from the BUSS cask. MICROSHIELD, which performs only gamma radiation shielding calculation, is based on a point-kernel model with idealized geometry, whereas SAS4 is a control module in the SCALE code system (1995) that can perform three-dimensional Monte Carlo shielding calculation for photons and neutrons, with built-in procedures for cross-section data processing and automated variance reduction. The two codes differ in how they model the details of the physics of gamma photon attenuation in materials, and this difference is reflected in the associated engineering cost of the analysis. One purpose of the analysis presented in this paper, therefore, is to examine the effects of the major modeling assumptions in the two codes on calculated dose rates, and to use the measured dose rates for comparison. The focus in this paper is on analysis of radiation dose rates measured on the general body of the cask and away from penetrations

  5. 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

  6. Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

    2006-01-01

    According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

  7. Effectiveness of shield materials in the design of the PFBR irradiated fuel subassembly shipping cask

    International Nuclear Information System (INIS)

    Radhakrishnan, G.

    2003-01-01

    Fuel subassemblies are irradiated inside the reactor core till they achieve the required burn up and after that they are cooled to permissible decay power level in in-vessel and ex-vessel storage places. Subsequently they are transported to reprocessing plants by means of shipping casks. Shield for the shipping cask has to be designed such a way that it has to comply with the ICRP recommended dose levels of less than 2 mSv/h on contact at the outer surface of the cask and less than 100 mSv/h at 1 m distance from the outer surface of the cask. In this paper, shield design of a typical PFBR irradiated fuel subassembly, which can transport three subassemblies at a time, is narrated. Considering the neutron and fission product and induced gamma rays emitted by typical PFBR irradiated core central subassembly subjected to a maximum burn up, as the source term shield design optimizations have been done. One-dimensional discrete ordinates transport theory computer code ANISN and point kernel computer code QAD-CGGP have been used in complement to carry out the shield design optimizations. Cast-iron, carbon steel, stainless steel 304 and lead and permali have been considered for shield materials. Shield requirements on top, bottom and along the axial height of the shipping cask have also been estimated. (author)

  8. 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

  9. Method for handling nuclear fuel casks

    International Nuclear Information System (INIS)

    Weems, S.J.

    1976-01-01

    A heavy shielded nuclear fuel cask is lowered into and removed from a water filled spent fuel pool by providing a vertical guide tube in the pool, affixing to the bottom of the cask a base plate that approximates the transverse dimension of the guide tube, and lowering and elevating the cask and base plate assembly into and out of the pool by causing it to traverse within the guide tube. The guide tube and base plate coact to function as a dashpot, thereby cushioning and controlling the fall of the cask in the pool should it break loose while being lowered into or raised out of the pool. a specified approach path to the guide tube insures that the cask assembly will not fall into the pool, should it break loose on its approach to the guide tube

  10. Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  11. Thermal analysis of the modified Hallum Nuclear Power Facility cask using experimentally obtained thermal boundary conditions corresponding to an engulfing open pool fire

    International Nuclear Information System (INIS)

    Longenbaugh, R.S.; Sanchez, L.C.; Gregory, J.J.

    1987-08-01

    This report presents the two-dimensional heat transfer analysis of an open pool fire surrounding a modified radioactive materials transport cask. The cask is an older cask that was used by the Hallum Nuclear Power Facility (HNPF). The HNPF cask did not have a neutron shielding region but was modified to include one for testing purposes. Analysis of the thermal effects of an engulfing open pool fire was performed with the use of the heat transfer code Q/TRAN, which had previously been used in thermal benchmarking problems for spent nuclear fuel casks. Boundary condition data for the analysis were derived from experimental open pool fire tests of large-scale calorimeter test articles performed at SNL that produced information about cask surface heat flux versus surface temperature relationships. Data analysis was directed toward a determination of the thermal response of the cask, particularly the extent of lead melt since lead is used within the HNPF cask's gamma-shielding region. Parameters, such as surface emissivity and internal heat generation rate, can affect the results of the thermal analysis which control the amount of lead melt. A parameter sensitivity analysis was performed using a one-dimensional model to describe how surface emissivity and internal heat generation rates affect the temperature distribution within the cask. The information from this analysis was used to determine the range of parameters for the two-dimensional thermal analysis. 13 refs., 57 figs., 8 tabs

  12. Safety analysis report for packaging: neutron shipping cask, model 4T

    International Nuclear Information System (INIS)

    Peterson, R.T.

    1977-01-01

    This Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual, Safety Standards for the Packaging of Fissile and Other Radioactive Materials

  13. XSDRNPM-S biasing of MORSE-SGC/S shipping-cask calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Tang, J.S.

    1982-06-01

    This report describes implementation of a systematic approach for biasing a Monte Carlo radiation transport calculation. In particular, the adjoint fluxes from a one-dimensional discrete ordinates calculation with the XSDRNPM-S code are used to generate biasing parameters for the multigroup Monte Carlo code, MORSE-SGC/S. Application of this biasing procedure to several deep penetration spent fuel shipping cask problems is also reported. The results obtained for neutron and gamma-ray transport indicate that relatively inexpensive Monte Carlo calculations are possible for dry and water filled shipping cask problems using these procedures. 5 tables

  14. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  15. 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

  16. Recommendations for cask features for robotic handling from the Advanced Handling Technology Project

    International Nuclear Information System (INIS)

    Drotning, W.

    1991-02-01

    This report describes the current status and recent progress in the Advanced Handling Technology Project (AHTP) initiated to explore the use of advanced robotic systems and handling technologies to perform automated cask handling operations at radioactive waste handling facilities, and to provide guidance to cask designers on the impact of robotic handling on cask design. Current AHTP tasks have developed system mock-ups to investigate robotic manipulation of impact limiters and cask tiedowns. In addition, cask uprighting and transport, using computer control of a bridge crane and robot, were performed to demonstrate the high speed cask transport operation possible under computer control. All of the current AHTP tasks involving manipulation of impact limiters and tiedowns require robotic operations using a torque wrench. To perform these operations, a pneumatic torque wrench and control system were integrated into the tool suite and control architecture of the gantry robot. The use of captured fasteners is briefly discussed as an area where alternative cask design preferences have resulted from the influence of guidance for robotic handling vs traditional operations experience. Specific robotic handling experiences with these system mock-ups highlight a number of continually recurring design principles: (1) robotic handling feasibility is improved by mechanical designs which emphasize operation with limited dexterity in constrained workspaces; (2) clearances, tolerances, and chamfers must allow for operations under actual conditions with consideration for misalignment and imprecise fixturing; (3) successful robotic handling is enhanced by including design detail in representations for model-based control; (4) robotic handling and overall quality assurance are improved by designs which eliminate the use of loose, disassembled parts. 8 refs., 15 figs

  17. Dynamic fracture toughness data for CASTOR registered casks

    International Nuclear Information System (INIS)

    Winkler, H.P.; Trubitz, P.; Pusch, G.; Warnke, E.P.; Beute, K.; Novotny, V.

    2004-01-01

    For the use of cast iron spherical graphite for heavy-sectioned casks for transportation and storage of radiactive materials a complete failure assessment including fracture mechanical analysis is necessary. The casks require an elaborate fracture mechanics design based on fracture mechanics evaluation. The extension of the existing code with respect to dynamic loading takes account new developments to extend the field of applications. It also includes new criteria to design these casks against operating and accident loadings. A fundamental requirement for the realisation of this standard and the calculation of admissible crack lengths of stresses under dynamic loads is the availability of fracture mechanical data. The paper presents-as a part of a large test-program-first results of dynamic fracture-toughness-investigations depending on structure and temperature. The test-program will incorporate investigations on more then 2500 specimens. The investigations that will be done include static and dynamic fracture mechanics tests, dynamic tensile and pressure-tests on different formed specimens. The temperatures and other test conditions follows the IAEA-regulations and the real service conditions. The test-program will be realised in partnership with different institutes

  18. Dynamic fracture toughness data for CASTOR {sup registered} casks

    Energy Technology Data Exchange (ETDEWEB)

    Winkler, H.P. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany); Trubitz, P.; Pusch, G. [Technische Univ. Bergakademie Freiberg, Freiberg (Germany); Warnke, E.P. [Siempelkamp GmbH and Co. KG, Krefeld (Germany); Beute, K. [Gontermann-Peipers GmbH, Siegen (Germany); Novotny, V. [SKODA, HUTE, Plzen (Czech Republic)

    2004-07-01

    For the use of cast iron spherical graphite for heavy-sectioned casks for transportation and storage of radiactive materials a complete failure assessment including fracture mechanical analysis is necessary. The casks require an elaborate fracture mechanics design based on fracture mechanics evaluation. The extension of the existing code with respect to dynamic loading takes account new developments to extend the field of applications. It also includes new criteria to design these casks against operating and accident loadings. A fundamental requirement for the realisation of this standard and the calculation of admissible crack lengths of stresses under dynamic loads is the availability of fracture mechanical data. The paper presents-as a part of a large test-program-first results of dynamic fracture-toughness-investigations depending on structure and temperature. The test-program will incorporate investigations on more then 2500 specimens. The investigations that will be done include static and dynamic fracture mechanics tests, dynamic tensile and pressure-tests on different formed specimens. The temperatures and other test conditions follows the IAEA-regulations and the real service conditions. The test-program will be realised in partnership with different institutes.

  19. Development of boronated aluminum alloy for basket of cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Sakaguchi, Y.; Saida, T.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities and competent authorities 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 cask uses new-developed boronated aluminum as basket material. This boronated aluminum has been developed to improve characteristics of material. To achieve this object, powder metallurgy method has been adopted for manufacturing boronated material. It is well known that this method provides excellent characteristics for the material and this boronated aluminum alloy has obtained excellent both mechanical and neutron absorbing characteristics. In addition, in order to maintain material properties for long-term use this boronated material is not strengthened by aging treatment. This paper summarizes an outline of the boronated aluminum alloy for basket assemblies by powder metallurgy. (author)

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

    International Nuclear Information System (INIS)

    Andress, D.; Joy, D.S.; McLeod, N.B.; Peterson, R.W.; Rahimi, M.

    1991-01-01

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

  1. 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.

  2. 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

  3. 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

  4. Safety analysis report: packages. Pu oxide and Am oxide shipping cask (Packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1980-05-01

    The PuO 2 cask or SP 5320-2 and 3 cask is designed for surface shipment of americium or plutonium. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, and Chapter I, Interstate Commerce Commission. The package has been assessed for transport of up to 357 grams of plutonium (403 grams PuO 2 powder) and up to 176 grams of americium (200 grams AmO 2 powder), having a maximum decay heat of 203 watts. Criticality evaluation alone would allow the shipment as Fissile Class II but the radiation level of the cask, measured at the time of shipment, may exceed 50 mrem/h at the surface and require shipment as Fissile Class III. Sample calculations address only the more restrictive of the two materials, which in most cases is 238 PuO 2

  5. T-3 cask users' manual. Revision 1

    International Nuclear Information System (INIS)

    1986-06-01

    This user's manual for the T-C spent fuel cask provides information on: operating procedures; inspection and maintenance procedures; criticality evaluation; shielding evaluation; thermal evaluation; structural evaluation; and limitations

  6. 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

  7. Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination

    Energy Technology Data Exchange (ETDEWEB)

    Bare, Walter Claude; Ebner, Matthias Anthony; Torgerson, Laurence Dale

    2001-08-01

    This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and Environmental Laboratory (INEEL) on a Gesellschaft für Nuklear Service (GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominion’s (formerly named Virginia Power) Surry Power Station and cask concrete storage pad. The assemblies have been continuously stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperatures, and cask surface gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket components including accessible weldments, fuel assembly exteriors, and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive, and mechanical examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the examination and testing indicate the concrete storage pad, CASTOR V/21 cask, and cask contents exhibited sound structural and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of gaseous fission products between 1985 and 1999.

  8. Safety analysis report: packages cobalt-60 shipping cask (packaging of radioactive and fissile materials)

    International Nuclear Information System (INIS)

    Evans, J.E.; Langhaar, J.W.

    1973-07-01

    Safety Analysis Report DPSPU-73-124-1 replaces DPSPU-69-124-1 and Supplement 1 to permit shipment of 350,000 curies of 60 Co (maximum) in cobalt-60 shipping casks in compliance with 10 CFR Part 71, Packaging of Radioactive Materials for Transport

  9. 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)

  10. GA-4/GA-9 legal weight truck from reactor spent fuel shipping casks

    International Nuclear Information System (INIS)

    1990-04-01

    The preliminary design report presents the results of General Atomics (GA) preliminary design effort to develop weight truck from reactor spent fuel shipping casks. The thermal evaluation of the Office of Civilian Radioactive Waste Management (OCRWM) cask considered normal and hypothetical accident conditions of transport. We employed analytical modeling as well as fire testing of the neutron shielding material to perform the evaluation. This document addresses the thermal design features of the cask, discusses thermal criteria, and summarizes the results of the thermal evaluation, as well as results of structural containment and nuclear evaluations that support the design. Also included are the results of trade-off studies. 69 refs., 103 figs., 76 tabs

  11. 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

  12. 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

  13. Transport of oxide spent fuel. Industrial experience of COGEMA

    International Nuclear Information System (INIS)

    Lenail, B.

    1983-01-01

    COGEMA is ruling all transports of spent fuel to La Hague reprocessing plant. The paper summarizes some aspects of the experience gained in this field (road, rail and sea transports) and describes the standards defined by COGEMA as regards transport casks. These standards are as follows: - casks of dry type, - casks of the maximum size compatible with rail transports, - capability to be unloaded with standardized equipment and following standard procedures

  14. Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Razvi, J.; Johnson, L.; Welch, B.; Lancaster, D.

    2004-01-01

    General Atomics (GA) has developed the Model GA-4 Legal Weight Truck Spent Fuel Cask, a high capacity cask for the transport of four PWR spent fuel assemblies, and obtained a Certificate of Compliance (CoC No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorized contents in this CoC however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorized contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burnup credit per ISG-8 Rev. 2, the authorized contents can be significantly expanded by increasing the maximum enrichment as the burnup increases. Use of burnup credit eliminates much of the criticality imposed limits on authorized package contents, but shielding still limits the use of the cask for the higher burnup, short cooled fuel. By downloading to two assemblies and using shielding inserts, even the high burnup fuel with reasonable cooling times can be transported

  15. 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.)

  16. 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.

  17. 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.)

  18. 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...

  19. Free drop impact analysis of shipping cask

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1989-01-01

    The WHAMS-2D and WHAMS-3D codes were used to analyze the dynamic response of the RAS/TREAT shielded shipping cask subjected to transient leadings for the purpose of assessing potential damage to the various components that comprise the the cask. The paper describes how these codes can be used to provide and intermediate level of detail between full three-dimensional finite element calculations and hand calculations which are cost effective for design purposes. Three free drops were adressed: (1) a thirty foot axial drop on either end; (2) a thirty foot oblique angle drop with the cask having several different orientations from the vertical with impact on the cask corner; and (3) a thirty foot side drop with simultaneous impact on the lifting trunnion and the bottom end. Results are presented for two models of the side and oblique angle drops; one model includes only the mass of the lapped sleeves of depleted uranium (DU) while the other includes the mass and stiffness of the DU. The results of the end drop analyses are given for models with and without imperfections in the cask. Comparison of the analysis to hand calculations and simplified analyses are given. (orig.)

  20. Criticality studies for dry storage cask

    International Nuclear Information System (INIS)

    Krishnani, P.D.; Srinivasan, K.R.

    1993-01-01

    Spent nuclear fuel from Tarapur Atomic Power Station (TAPS) is stored in a storage pool located inside the reactor building. The capacity of this pool was initially to meet storage requirements of 528 bundles which was later augmented from time to time. Since the enhanced capacity was also getting exhausted, setting up of a storage pool away from reactor was envisaged. As an interim measure, the dry storage casks were designed to store the spent fuel already cooled for a few years in the storage pools. If water enters the cask, the cask interior may be covered with steam water or air-water mixture. This paper gives the results of criticality calculations for storage cask under various conditions of steam water mixture, using the computer code LWRBOX. In these calculations, it has been assumed that the cask contains the most reactive fuel assemblies of reload-1 at zero burnup. It also gives the comparison of some of the results with General Electric (GE) calculations. (author). 3 refs., 1 fig., 2 tabs

  1. Status of radiation shield design for liquid metal fast breeder reactor spent fuel shipping cask application

    International Nuclear Information System (INIS)

    Dupree, S.A.; Rack, H.J.

    1976-09-01

    Neutron and gamma-ray transport calculations in one-dimensional cylindrical geometry have been performed on a trial reference LMFBR spent-fuel shipping cask that could transport one CRBR subassembly. In the study it was assumed that a layer of depleted U and a layer of neutron shielding materials were sandwiched between 5.08-cm-thick (2-in.) layers of stainless steel. The thicknesses of the internal layers were adjusted until a balanced dose rate (50 percent neuton and 50 percent gamma-ray) of 5 mrem/hr was achieved at a point 1.83 m (6 ft) from the cask surface. Neutron-shield materials considered were LiH, Be, B 4 C, DiH 2 . 5 , and C (graphite). Of these materials, LiH provided the smallest, lightest, and least expensive cask; however, its use would be contigent on expansion of production facilities for LiH and development of a canning or cladding procedure. The B 4 C shielded cask would offer the best alternative if the designs were limited to those using currently available materials

  2. Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

    International Nuclear Information System (INIS)

    Wagner, John C.

    2003-01-01

    This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)

  3. A fuel response model for the design of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Duffey, T.A.; Einziger, R.E.; Hobbins, R.R.; Jordon, H.; Rashid, Y.R.; Barrett, P.R.; Sanders, T.L.

    1989-01-01

    The radiological source terms pertinent to spent fuel shipping cask safety assessments are of three distinct origins. One of these concerns residual contamination within the cask due to handling operations and previous shipments. A second is associated with debris (''crud'') that had been deposited on the fuel rods in the course of reactor operation, and a third involves the radioactive material contained within the rods. Although the lattermost source of radiotoxic material overwhelms the others in terms of inventory, its release into the shipping cask, and thence into the biosphere, requires the breach of an additional release barrier, viz., the fuel rod cladding. Hence, except for the special case involving the transport of fuel rods containing previously breached claddings, considerations of the source terms due to material contained in the fuel rods are complicated by the need to address the likelihood of fuel cladding failure during transport. The purpose of this report is to describe a methodology for estimating the shipping cask source terms contribution due to radioactive material contained within the spent fuel rods. Thus, the probability of fuel cladding failure as well as radioactivity release is addressed. 8 refs., 2 tabs

  4. 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...

  5. Radiation Templates of Spent Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Vanier, Peter

    2018-05-07

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

  6. Spent and fresh fuel shipping cask considerations

    International Nuclear Information System (INIS)

    Shappert, L.B.; Unger, W.E.; Freedman, J.M.

    1975-01-01

    A program to provide basic information for cask design and safety has been conducted for over ten years at Oak Ridge National Laboratory. Principal problem areas in Liquid Metal Fast Breeder Reactor (LMFBR) casks are identified as heat transfer, structures and containment, criticality and shielding. Solutions in the problem areas, as well as the need for future work, are addressed by describing an LMFBR conceptual design cask. A new program, which is underway at Sandia Laboratories, Albuquerque, New Mexico, is aimed at producing technology useful to industry and government. Technologies are being developed in areas of hazards analysis, heat transfer, shielding, structures and containment, and spent fuel characterization, substantiated by hot laboratory verification. Particular emphasis will be placed on establishing qualification tests based on accident experience. Handling requirements and limitations are discussed. (auth)

  7. Research and development of spent-fuel shipping casks and the criteria for sea-going vessels carrying them

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Since the transport of spent fuel will increase rapidly and extensively in the near future, the Japanese Atomic Energy Committee enacted the Technical Standard for Transportation of Radioactive Materials, based on the IAEA Regulation for the Safe Transport of Radioactive Materials, 1973 Revised Edition. The authorities concerned have begun to review the former ordinances for transporting radioactive materials and to develop a unified system of relevant laws and standards. For ten years the Atomic Energy Bureau has invested in research and development to obtain data for the design and licence of a spent-fuel shipping cask. Different scale models of a prototype weighing 80t were used to clarify the scale effect of drop, puncture and fire tests, which are a feature of Japanese research and development. Also an immersion test in water at pressures up to about 500 bar is now carried out to investigate the integrity of the cask body and sealing structure to prevent leakage of radioactive contents to the surroundings should the cask fall into deep sea. In Japan, depending on the site of nuclear plants, almost all transport of unirradiated and spent fuels is by sea. Therefore, to secure safe transport, the design criteria of ships for the exclusive transport of spent-fuel shipping casks, namely full-load shipping, have been enacted, which aim to make the likelihood of sinking on collision, stranding, and other unforeseen accidents at sea highly improbable and also to keep radiation exposure of the crew as low as possible. (author)

  8. Developments in shielding and criticality assessment for cask design

    International Nuclear Information System (INIS)

    Watmough, M.H.; Cooper, A.J.; Croxford, R.W.

    1993-01-01

    This paper presents recent highlights from the shielding and criticality methods development programme that are of relevance to cask design. Specifically, the following points emerge: 1) the preparation of a licence application based upon UK methods and data used in a standardized fissile depletion and plutonium production model has been completed; 2) the assumptions used in the modelling of granules of broken fuel within the transport package following a postulated impact accident have been revised thereby allowing less pessimistic assessments to be performed; 3) enhancements are being made to the software used for shielding and criticality analysis enabling a more cost effective design service to be provided. These ongoing developments clearly show the activity to extend the scope of assessments while increasing the physical realism of the models. Through these developments BNFL continues to offer a comprehensive and cost effective shielding and criticality analysis service as part of its worldwide fuel transport business. (J.P.N.)

  9. Marginal overweight operating scenario for DOE's initiative I highway casks

    International Nuclear Information System (INIS)

    Hill, C.V.; Loud, G.C.; Heitzman, A.C.

    1993-01-01

    This paper assesses the potential transport of high-capacity Initiative I highway casks under development by the Office of Civilian Radioactive Waste Management (OCRWM) as permitted marginal overweight shipments that: exceed a gross vehicle weight (gvw) limit of 80,000, but weight less than 96,000 pounds; follow axle and axle group weight limits adopted by the Surface Transportation Assistance Act (STAA) of 1982; conform to dimensional restrictions to operate on most major highways; and comply with the Federal Bridge Formula. The marginal overweight tractor-trailer would operate in normal open-quotes over-the-roadclose quotes mode and comply with all laws and regulations. The vehicle would have a sleeper berth and two drivers - one to drive while the other provides escort and communications services and accumulates required off-duty time

  10. 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.)

  11. 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).

  12. Transfer cask system design activities: status and plan

    International Nuclear Information System (INIS)

    Locke, D.; Gutierrez, C. Gonzalez; Damiani, C.; Gracia, V.; Friconneau, J.-P.; Martins, J.-P.; Blight, J.

    2011-01-01

    The ITER Cask and Plug Remote Handling System (CPRHS), a.k.a. Transfer Cask System, is a critical element of the ITER Remote Maintenance System (IRMS) devoted to transportation of components between the Tokamak building and Hot Cell. Due to the necessary confinement of contaminated components the CPRHS is defined as Safety Importance Class 1 (SIC-1) plus the mobile nature of the CPRHS brings with it a significant number of complex interfaces with other ITER sub-systems. With a total CPRHS fleet in excess of 20 units, including seven typologies, the management of design and procurement needs to be carefully planned and implemented to ensure compliance with ITER's requirements. Fusion for Energy (F4E) and its beneficiaries/contractors are currently working under ITER Task Agreements (ITAs) on the conceptual design of the CPRHS and, following the signing of the Procurement Arrangement (PA) in mid 2012, will take responsibility for the entire CPRHS fleet. F4E must, therefore, develop a robust strategy to meet the needs of both ITER machine assembly (for which a number of CPRHS units will be utilised) and the remote maintenance of ITER. Within this context this paper will present the status of the current CPRHS design activities, highlight some of the significant issues which will be faced during procurement and present the overall strategy which is being implemented by F4E in order to meet these challenging objectives.

  13. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Gartz, R.; Kuehne, B.; Diersch, R.

    2004-01-01

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

  14. 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

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

    International Nuclear Information System (INIS)

    Chen, T.F.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.; Mok, G.C.

    1995-02-01

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

  16. A revision of the Cask Designers Guide for the '90s

    International Nuclear Information System (INIS)

    Shappert, L.B.; Green, V.M.

    1992-01-01

    The report A Guide for the Design Fabrication, and Operation of Shipping Casks for Nuclear Applications, ORNL-NSIC-68, commonly called the Cask Designers Guide, is being revised at the request of the Transportation and Packaging Safety Division of the Department of Energy (DOE). The new document will be called the Packaging Handbook. The Cask Designers Guide was published in 1970 during the period when many radioactive materials packagings were being developed and many technical studies applicable to these packagings were being performed. Since that period, many improvements in packaging design have appeared, designers have improved their calculational techniques, and much effort has gone into applying Quality Assurance (QA) principles to cask development Materials, and their limitations, have surfaced as a very important consideration in the licensing process. While the Packaging Handbook considers all Type B packages, most of the authors' experience lies in the technical areas found in the licensing of spent nuclear fuel (SNF) packagings and this is reflected in the document

  17. Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

    International Nuclear Information System (INIS)

    Apostolov, T; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

  18. Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Buyun, E-mail: ayun@iim.ac.cn [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Gao, Lifu [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Cao, Huibin; Sun, Jian [Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Sun, Yuxiang; Song, Quanjun; Ma, Chengxue; Chang, Li; Shuang, Feng [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2014-12-15

    Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.

  19. Structural dimensioning of dual purpose cask prototype

    International Nuclear Information System (INIS)

    Silva, Luiz Leite da; Mourao, Rogerio Pimenta; Lopes, Claudio Cunha

    2005-01-01

    The structural dimensioning of a Type B(U) dual purpose cask prototype is part of the scope of work of the Brazilian institute CDTN in the IAEA regional project involving Latin American countries which operate research reactors (Argentina, Brazil, Chile, Mexico and Peru). In order to meet the dimensional and operational characteristics of the reactor facilities in these countries, a maximum weight of 10.000 kgf and a maximum dimension of 1 m in at least one direction were set for the cask. With these design restrictions, the cask's payload is either 21 MTR or 78 TRIGA fuel elements. The cask's most important components are main body, primary and secondary lids, basket and impact limiters. The main body has a sandwich-like wall with internal and external layers made of AISI 304 stainless steel with lead in-between. The lead provides biological shielding. The primary lid is similarly layered, but in the axial direction. It is provided with a double system of metallic rings and has ports for pressurization, sampling and containment verification. The secondary lid has the main function of protecting the primary lid against mechanical impacts. The basket structure is basically a tube array reinforced by bottom plate, feet and spacers. Square tubes are used for MTR elements and circular tubes for TRIGA elements. Finally, the impact limiters are structures made of an external stainless steel thin covering and a filling made of the wood composite OSB - Oriented Strand Board. The prototype is provided with bottom and top impact limiters, which are attached to each other by means of four threaded rods. The limiters are not rigidly attached to the cask body. A half scale cask model was designed to be submitted to a testing program. As its volume scales down to 1:8, the model weight is 1,250 kgf. This paper presents the methodology for the preliminary structural dimensioning of the critical parameters of the cask prototype. Both normal conditions of operation and hypothetical

  20. 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)

  1. 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

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

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  4. Ductile iron cask with encapsulated uranium, tungsten or other dense metal shielding

    International Nuclear Information System (INIS)

    Barnhart, V.J.; Anderson, R.T.

    1989-01-01

    In a cask for the transportation and storage of radioactive materials, an improvement in the shielding means which achieves significant savings in weight and increases in payload by the use of pipes of depleted uranium, tungsten or other dense metal, encapsulating polyethylene cores, dispersed in two to four rows of concentric boreholes around the periphery of the cask body which is preferably made of ductile iron. Alternatively, rods or small balls of these same shielding materials, alone or in combination, are placed in these bore holes. The thickness, number and arrangement of these shielding pipes or rods is varied to provide optimum protection against the neutrons and gamma radiation emitted by the particular radioactive material being transported or stored. (author) 4 figs

  5. Resin composites

    DEFF Research Database (Denmark)

    Benetti, Ana Raquel; Peutzfeldt, Anne; Lussi, Adrian

    2014-01-01

    OBJECTIVE: To investigate how the modulus of elasticity of resin composites influences marginal quality in restorations submitted to thermocyclic and mechanical loading. METHODS: Charisma, Filtek Supreme XTE and Grandio were selected as they were found to possess different moduli of elasticity...... of resin composite (p=0.81) on the quality of dentine margins was observed, before or after loading. Deterioration of all margins was evident after loading (p....008). CONCLUSIONS: The resin composite with the highest modulus of elasticity resulted in the highest number of gap-free enamel margins but with an increased incidence of paramarginal enamel fractures. CLINICAL SIGNIFICANCE: The results from this study suggest that the marginal quality of restorations can...

  6. 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.

  7. Thermo-mechanical finite element analyses of bolted cask lid structures

    International Nuclear Information System (INIS)

    Wieser, G.; Qiao Linan; Eberle, A.; Voelzke, H.

    2004-01-01

    The analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak-tightness in package design assessment according to the Transport Regulations or in aircraft crash scenarios. In this context BAM is developing methods based on Finite Elements to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. I n case of fire it might be not enough to perform only a thermal heat transfer analysis. The complex cask design in connection with a severe hypothetical time-temperature-curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and its counterparts that can be analyzed by a so-called thermo-mechanical calculation. Although it is currently not possible to correlate leakage rates with results from deformation analyses directly an appropriate Finite Element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field respectively. Except of the lid bolts the geometry of the cask and the mechanical loading is axial-symmetric which simplifies the analysis considerably and a two-dimensional Finite Element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modelling on the relative displacements at the seating of the seals. Besides this, the influence of bolt modelling, thermal properties and detail in geometry of the two-dimensional Finite Element models on

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  9. Half and full scale drop tests for qualification of CONSTOR registered casks as type B(U)F packages

    International Nuclear Information System (INIS)

    Koenig, S.; Diersch, R.

    2004-01-01

    The CONSTOR registered steel-sandwich cask was developed as a cost-effective design by using conventional mechanical engineering technologies and commonly available materials. The CONSTOR registered consists of a cask body with an outer and an inner shell made of steel. At the upper end, the shells are welded to a ring made of forged steel. The space between the two shells is filled with heavy concrete for gamma and neutron shielding. The liner of the casks and the forged head ring form the basis for the structural integrity, the concrete bears only a menial part of accident loads. The CONSTOR registered concept fulfils both the internationally valid IAEA criteria for transportation and the requirements for long-term intermediate storage in the US and various European countries. Since the beginning of the development of the CONSTOR registered design, two drop test series have already been performed and a third one is being planned to start in 2004

  10. Thermal analyses of the IF-300 shipping cask

    International Nuclear Information System (INIS)

    Meier, J.K.

    1978-07-01

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

  11. 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

  12. Conceptual evaluation of type B(U) casks for the nuclear power plants of Argentina

    International Nuclear Information System (INIS)

    Florido, P.C.; Isnardi, E.R.

    1993-01-01

    In Argentina two different nuclear power plants are in operation, Atucha I (PHWR-Siemens) and Embalse (PHWR-CANDU). Thus two very different fuel elements could be potentially transported. In order to optimize the research and development needed for the design and construction of the cask, the cost-benefit and flexibility of the engineering solutions are studied, for the two fuel elements. Different casks, for both types of existing fuel elements (Atucha I and Embalse), for different burnup-levels (regarding the advanced fuel cycle available), decay times, distances, and transported weight were studied. Three materials for shielding were used: uranium lead and steel. Only transport by road was considered, due to the reduced availability of the train. In this stage (conceptual design) small, easy and fast computer programs should be used. The principal issues that have to be fixed are shielding properties, thermal effects and mechanical behavior. As a result of the evaluation, different options for the casks were founded, as well as the importance of different parameters and the effect of two different designs of fuel elements. (author)

  13. 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)

  14. 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

  15. 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)

  16. Rail tiedown tests with heavy casks for radioactive shipments

    International Nuclear Information System (INIS)

    Petry, S.F.

    1980-08-01

    A rail tiedown test program was conducted at the Savannah River Plant in July and August 1978. For each test, a 40- or 70-ton cask was secured on a railcar. The railcar was pushed to speeds up to 11 mph and allowed to couple to parked railcars simulating ordinary railyard operations. The test car carrying the cask was heavily instrumented to measure the accelerations and forces generated at strategically selected places. Eighteen test runs were made with different combinations of railcars, couplers, casks, speeds, and tiedown configurations. The major objectives of the test program were to (1) provide test data as a basis to develop a tiedown standard for rail cask shipments of radioactive materials and (2) collect dynamic data to support analytical models of the railcar cask tiedown system. The optimum tiedown configuration demonstrated for heavy casks was a combination of welded, fixed stops to secure the cask longitudinally and flexible cables to restrain vertical and lateral cask movement. Cables alone were inadequate to secure a heavy cask to a standard railcar, and bolting was found disadvantageous in several respects. The use of cushioning coupler mechanisms dramatically reduced the tiedown requirements for the rail coupling operation. The test program and general conclusions are discussed

  17. Closure for casks containing radioactive materials

    International Nuclear Information System (INIS)

    Hall, G.V.B.; Mallory, C.W.

    1990-01-01

    This patent describes an improved closure for covering and sealing an opening in a single cask for containing radioactive material, wherein the opening is characterized by a ledge. It comprises: an inner lid receivable within the opening and having a gasket means that is seatable over the ledge; an outer lid which is likewise receivable into the opening and securable therearound when the outer lid is rotated relative to the opening. The inner lid remaining stationary relative to the cask opening when the outer lid is rotated and having no torque applied thereto by the outer lid when the outer lid is rotated, and bolt means threadedly mounted through the outer lid for applying a compressive force between the inner and outer lids after the outer lid has been secured to the opening in order to depress the gasket means of the inner lid into sealing engagement with the ledge while avoiding the application of torsion between the gasket means and the ledge

  18. MCO loading and cask loadout technical manual

    International Nuclear Information System (INIS)

    PRAGA, A.N.

    1998-01-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description

  19. The TN-RAM - a new cask for shipping high activity irradiated hardware

    International Nuclear Information System (INIS)

    Neider, T.; Hanson, A.S.

    1993-01-01

    Transnuclear, Inc. has developed a Type B(U) radioactive material transport packaging designed specifically for the transport of dry, irradiated, non-fuel bearing components (NFBC). The TN-RAM is a transport cask in the configuration of a right circular cylinder, fabricated from lead and stainless steel, with wood-filled impact limiters attached at both ends. The lead and stainless steel construction of the lid, walls, and bottom provides a shielding effectiveness of approximately 7.1 inches (18 cm) lead equivalent. (J.P.N.)

  20. Spray drying of bead resins: feasibility tests

    International Nuclear Information System (INIS)

    Gay, R.L.; Grantham, L.F.; Jones, L.J.

    1984-01-01

    Rockwell International has developed a volume reduction system for low-level reactor wastes based on drying the wastes in a heated-air spray dryer. The drying of slurries of sodium sulfate, boric acid, and powdered ion exchange resins was demonstrated in previous tests. The drying of bead ion exchange resins can be especially difficult due to the relatively large size of bead resins (about 500 to 800 microns) and their natural affinity for water. This water becomes part of the pore structure of the resins and normally comprises 50 t 60 wt % of the resin weight. A 76-cm-diameter spray dryer was used for feasibility tests of spray drying of cation and anion bead resins. These resins were fed to the dryer in the as-received form (similar to dewatered resins) and as slurries. A dry, free-flowing product was produced in all the tests. The volume of the spray-dried product was one-half to one-third the volume of the as-received material. An economic analysis was made of the potential cost savings that can be achieved using the Rockwel spray dryer system. In-plant costs, transportation costs, and burial costs of spray-dried resins were compared to similar costs for disposal of dewatered resins. A typical utility producing 170 m 3 (6,000 ft 3 ) per year of dewatered resins can save $600,000 to $700,000 per year using this volume reduction system

  1. Behaviour of neutron moderator materials at high temperatures in CASTOR registered -casks: qualification and assessment

    International Nuclear Information System (INIS)

    Krietsch, T.; Wolff, D.; Knopp, U.; Brocke, H.D.

    2004-01-01

    The Federal Institute for Materials Research and Testing (BAM) is the responsible German authority for the assessment of mechanical and thermal designs of transport and storage casks for radioactive materials. BAM checks up the proofs of the applicants in their safety reports and assesses the conformity to the Regulations for the Safe Transport of Radioactive Material. One applicant is the Gesellschaft fuer Nuklear-Behaelter mbH (GNB) with a new generation of transport and storage casks of CASTOR registered -design. GNB typically uses ultra high molecular weight Polyethylene (UHMW-PE) for the moderation of free neutrons. Rods made of UHMW-PE are positioned in axial bore holes in the wall of the cask and plates of UHMW-PE are in free spaces between primary and secondary lid and between the bottom of the cask and an outer plate (Figure 1). Because of the heat generated by the radioactive inventory and because of a strained spring at the bottom of every bore hole, UHMW-PE is subjected to permanent thermal and mechanical loads as well as loads from gamma and neutron radiation. UHMW-PE has been used under routine- and normal conditions of transport for maximum temperatures up to 130 C. For new generations of CASTOR registered -design maximum temperatures will be increased up to 160 C. That means a permanent use of UHMW-PE at temperatures within and above the melting region of the crystallites. In this paper, some results of special investigations for the proofs of usability of UHMW-PE at temperatures up to 160 C under real conditions of transport and storage in CASTOR registered -casks are given. For that, investigations on temperature dependent expansion behaviour under laboratory conditions as well as in large scale experiments, especially in the case of multiple heating and cooling, were done. Besides, geometrical creep strength for long-term loading by temperatures and pressures with regard to the chemical and physical stability properties of UHMW-PE above the

  2. A conceptual redesign of an Inter-Building Fuel Transfer Cask

    International Nuclear Information System (INIS)

    Klann, R.T.; Picker, B.A. Jr.

    1993-01-01

    The Inter-Building Fuel Transfer Cask, referred to as the IBC, is a lead shielded cask for transporting subassemblies between buildings on the Argonne National Laboratory-West site near Idaho Falls, Idaho. The cask transports both newly fabricated and spent reactor subassemblies between the Experimental Breeder Reactor-II (EBR-II), the Fuel Cycle Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The IBC will play a key role in the Integral Fast Reactor (IFR) fuel recycling demonstration project. This report discusses a conceptual redesign of the IBC which has been performed. The objective of the conceptual design was to increase the passive heat removal capabilities, reduce the personnel radiation exposure and incorporate enhanced safety features into the design. The heat transfer, radiation and thermal-hydraulic properties of the IBC were analytically modelled to determine the principal factors controlling the desip. The scoping studies that were performed determined the vital physical characteristics (i.e., size, shielding, pumps, etc.) of the MC conceptual design

  3. QA in the design and fabrication of the TMI-2 rail cask

    International Nuclear Information System (INIS)

    Hayes, G.R.

    1988-01-01

    EGandG Idaho, Inc., acting on behalf of the US Department of Energy, is responsible for transporting core debris from Three Mile Island-Unit 2 to the Idaho National Engineering Laboratory. Transportation of the debris is being accomplished using an NRC licensed container, called the NuPac 125-B. This paper describes the NuPac 125-B Rail Cask and the quality assurance (QA) requirements for that system. Also discussed are the QA roles of the various organizations involved in designing, building, inspecting and testing the NuPac 125-B. The paper presents QA/QC systems implemented during the design, procurement, and fabrication of the cask to assure compliance with all applicable technical codes, standards and regulations. It also goes beyond the requirements aspect and describes unique QA/QC measures employed to assure that the cask was built with minimum QA problems. Finally, the lessons learned from the NuPac 125-B project is discussed. 4 refs., 4 figs

  4. Finite element analysis for the impact behaviour of a cask interacting with a rigid pin

    International Nuclear Information System (INIS)

    Altes, J.; Geiser, H.; Voelzer, W.; Frenk, A.; Deeken, G.

    1993-01-01

    Full scale drop tests of casks to be licensed as type B packages according to the IAEA regulations for the safe transport of radioactive materials are expensive. Therefore efforts are being made to use computer codes for calculating the impact behaviour. But these codes have to be verified by experiments. Codes available for these calculations are for example DYNA3D and ABAQUS. In the paper results of both codes are compared. A 11 t ductile cast iron cask (type MOSAIK) without impact limiters was analysed dropping from a height of 1 m with its top onto a cylindrical steel pin. The results of the finite element calculations with both codes show good agreement. The ABAQUS results using the implicit method are in accordance with the explicit method, for which considerably shorter CPU times are noted. (author)

  5. 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)

  6. 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.)

  7. The role of ORIGEN-S in the design of burnup credit spent fuel casks

    International Nuclear Information System (INIS)

    Brady, M.C.

    1991-01-01

    Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative ''fresh fuel assumption'' be used in the criticality analysis. Burnup credit refers to a new approach in criticality analyses for spent fuel handling systems in which reactivity credit is allowed for the depleted state of the fuel. Studies have shown that the increased cask capacities that can be achieved with burnup credit offer both economic and risk incentives. The US Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. Specifically, through the Shielding Analysis Sequence 2H (SAS2H), ORIGEN-S is linked with cross-section processing codes and one-dimensional transport analyses to produce problem-specific cross-section data for the point-depletion calculation. The utility code COUPLE facilitates updating basic cross-section and fission-yield data for the calculations. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process

  8. Nondestructive evaluation of LWR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined

  9. 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.)

  10. CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding

    International Nuclear Information System (INIS)

    1988-01-01

    1 - Description of program or function: CAPSIZE is an interactive program to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel casks designed to meet those objectives. 2 - Method of solution: Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the load cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium- shielded cask meeting those objectives. Using optimal packing arrangements and shielding requirements input by the user, SCOPE will design a cask to carry a single fuel assembly and then continue incrementing the number of assemblies until one or more of the design limits can no longer be met. KWIKDOSE queries the user for the number of PWR fuel assemblies in a cask, the type of cask and thickness of the shield. Upon getting the necessary input, KWIKDOSE prints out the total dose rate, 10 feet from the centerline of the cask, as a function of the burnup and cooling time of the spent fuel. 3 - Restrictions on the complexity of the problem: The restrictions are subject to the shielding requirements of the shipping cask

  11. Application of the tack weld to cask impact limiter case

    International Nuclear Information System (INIS)

    Ku, J. H.; Choung, W. M.; You, G. S.; Park, S. W.

    2001-01-01

    The objective of this paper is to evaluate the benefit of the application of intermittent tack weld to the cask impact limiter case in the cask impact accident. This paper describes the test results of weldment rupture of foam filled tube type energy absorber and analytical evaluation of the effect of intermittent tack weld to the cask impact limiter case on the cask impact behavior. Prior to the cask impact analysis, the evaluation of weldment joint was carried out for intermittent tack weldment considering the weldment rupture. The intermittent tack welded part is weaker than ordinary weldment so ruptured if the stress exceeds certain limit. The rupture of the impact limiter case causes to lose its constraining effect for the wood blocks, which are filled into the metal incasement between the case and the gussets. The application of intermittent tack weld to the impact limiter case showed great advantage in vertical and horizontal drop impacts

  12. Cask crush pad analysis using detailed and simplified analysis methods

    International Nuclear Information System (INIS)

    Uldrich, E.D.; Hawkes, B.D.

    1997-01-01

    A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped spent nuclear fuel shipping cask into a 44-ft. deep cask unloading pool at the Fluorinel and Storage Facility (FAST). This facility, located at the Idaho Chemical Processing Plant (ICPP) at the Idaho national Engineering and Environmental Laboratory (INEEL), is a US Department of Energy site. The basis for this study is an analysis by Uldrich and Hawkes. The purpose of this analysis was to evaluate various hypothetical cask drop orientations to ensure that the crush pad design was adequate and the cask deceleration at impact was less than 100 g. It is demonstrated herein that a large spent fuel shipping cask, when dropped onto a foam crush pad, can be analyzed by either hand methods or by sophisticated dynamic finite element analysis using computer codes such as ABAQUS. Results from the two methods are compared to evaluate accuracy of the simplified hand analysis approach

  13. 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