[en] 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

[en] This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area

[en] 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

[en] The Swedish KBS-3 design for the disposal of spent fuel is based on encapsulation of the fuel in canisters consisting of cast iron inserts and an outer 5 cm thick shield of copper. The canisters are embedded in bentonite clay and will be disposed in crystalline bedrock at a depth of about 500 m. To verify that the canisters fulfil the requirements, an extensive program for quality control is developed. In this program the use of NDT is vital and therefore it is very important to determine the reliability of the developed NDT methods. The reliability of the NDT methods has been studied for six years and the results are used in several ways. One task is to identify areas where further optimizations of the NDT methods are necessary and another is to study the effect of influencing parameters prior to the upcoming qualifications. Furthermore, the reliability analyses of the NDT are combined with analyses of the manufacturing and welding processes to give input to the overall safety analyses of the long-term properties of the canisters. (orig.)

[en] The aim of this paper is to determine the main physical quantities of the spent fuel casks from the research reactor IRT-2000 Sofia, which are to be transported. The methodology for determination of the isotope inventory and the residual energy release is based on the program complex NESSEL-MIKRO-NUKO. Using this product it is determined the isotope inventory, residual power of the all 73 spent fuel casks of type EK-10 and C36 for the operation period 1961-1989 and the storage period after the reactor shutdown in 2000. The following data is prepared for each cask: regime of operation including working hours, outage, and produced energy; mass of the Uranium isotopes and ²³⁵U of the fresh fuel; combustion depth; combustion rate; residual power release; isotope composition; mass and gamma activity of the isotopes

[en] Fisher-Helium Purge Flow Control Valve and Relay

[en] 'TN international' has developed a capacity to envision and realize innovative products and services for the transport and storage of radioactive materials. This is our innovation policy, whose aim is managing change by design. The paper describes a new concept, internally known as ID school. It is a unique combination of innovation tools and methods. It blends an all-round technological watch, a idea management system, an extensive research/development program, state-of-the-art research and problem-solving software, a network of experts, an array of innovation methods and facilities that promote an innovative spirit. These tools apply to engineering, freight forwarding, organization... and favour both individual initiative and teamwork. For instance, AREVA has developed a special method for an innovation called the EFICA method. This method alternates diverging and converging phases to conclude with a set of innovative ideas and an action plan to develop them. In this method, brainstorming is stimulated by facilitators. These are engineers who receive a specific training. More than twenty EFICA projects have been carried out very successfully by TN International, allowing many new ideas to emerge. These ideas are patented and implemented in cask designs. Examples are shown of success coming from innovation challenges: more capacious packaging, accommodation of hotter materials, reduction of dose rates, quicker approvals. The expected benefits are: -) shorter time-to-market, -) cheaper products, -) differentiated products, and -) safer products. Moreover, our experience of innovation initiatives show that the challenges from Safety Authorities can also be met: increased safety, thorough justifications, implementation of good existing ideas to new problems. Therefore, innovation can be looked upon favourably by competent authorities

[en] Cosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons. (authors)

[en] To provide a cask with the largest possible loading capacity of spent fuel assemblies with the largest practicable burnup and shortest cooling time within all safety requirements, AREVA TN has adapted its design process and developed a more elaborated shielding analysis method. AREVA TN reconsiders the standard definition of the content in order to take advantage of potential heterogeneities between sources of the loaded fuel assemblies and of the self-shielding between the loaded fuel assemblies. The maximum authorized radioactive content is defined with only maximum neutron and gamma sources authorized in each basket lodgement. The result of this method is expressed under the shape of a linear inequalities system allowing to optimize the cask capacity and performance. The linear inequalities system assures that the radiation level limits are respected and presents a high implementation flexibility and allows us to take advantage of the explicit characteristics of the fuel assembly inventory. This method avoids the necessity of defining in the license a maximal burnup and a minimum cooling time, authorized in order to respect the radiation level limits and it does not require any dose rate calculation for each loading

[en] The TN-MTR cask has been designed for the transport of spent fuel issuing from material testing reactors (MTR) or TRIGA reactors. This type B packaging will be allowed to be used in France and abroad. 3 types of fuel basket are planned, the biggest capacity basket is made up of 68 87.2*87.2 mm compartments and 8 smaller ones. The safety-criticality studies have proved that K_eff + 3s < 0.95 even in accidental conditions. A series of gravity drop tests has been led in march and april 1997, it was shown that the cask suffered only local deformations and that the mechanical resistance of the basket was sufficient. 3 TN-MTR casks have been manufactured and will be soon put into service. (A.C.)