[en] This system design description (SDD) addresses the Vacuum Purge System Chilled Water (VPSCHW) system. The discussion that follows is limited to the VPSCHW system and its interfaces with associated systems. The reader's attention is directed to Drawings H-1-82162, Cold Vacuum Drying Facility Process Equipment Skid PandID Vacuum System, and H-1-82224, Cold Vacuum Drying Facility Mechanical Utilities Process Chilled Water PandID. Figure 1-1 shows the location and equipment arrangement for the VPSCHW system. The VPSCHW system provides chilled water to the Vacuum Purge System (VPS). The chilled water provides the ability to condense water from the multi-canister overpack (MCO) outlet gases during the MCO vacuum and purge cycles. By condensing water from the MCO purge gas, the VPS can assist in drying the contents of the MCO

[en] CASTOR casks are designed for the transport and dry storage of spent light water reactor fuel or nuclear waste, like for instance vitrified waste. The subcriticality of CASTOR V casks is mainly assured by: the limitation of the fissile content of the fuel, the geometrical positioning of the fuel assemblies within a basket, and the neutron absorbing structures. Using the example of CASTOR V type casks, an outline of the methodology of criticality analyses for CASTOR casks is presented. The development of models for normal and accidental conditions of transport is presented. The criticality safety analyses are performed with the KENO codes and with the SCALE 238-group ENDF/B-V cross section library. This methodology has been developed over the last 15 years and has been successfully applied to various licensing procedures

[en] A crush pad has been designed and analyzed to absorb the kinetic energy of an accidentally dropped spent nuclear fuel shipping cask into a 44 ft. deep cask unloading pool. Conventional analysis techniques available for evaluating a cask for tipping due to lateral seismic forces assume that the cask rests on a rigid surface. In this analysis, the cask (110 tons) sits on a stainless steel encased (0.25 in. top plate), polyurethane foam (4 ft. thick) crush pad. As the cask tends to rock due to horizontal seismic forces, the contact area between the cask and the crush pad is reduced, increasing the bearing stress, and causing the pivoting corner of the cask to depress into the crush pad. As the crush pad depresses under the cask corner, the pivot point shifts from the corner toward the cask center, which facilitates rocking and potential tipping of the cask. Subsequent rocking of the cask may deepen the depression, further contributing to the likelihood of cask tip over. However, as the depression is created, the crush pad is absorbing energy from the rocking cask. Potential tip over of the cask was evaluated by performing a non-linear, dynamic, finite element analysis with acceleration time history input. This time history analysis captured the effect of a deforming crush pad, and also eliminated conservatisms of the conventional approaches. For comparison purposes, this analysis was also performed with the cask sitting on a solid stainless steel crush pad. Results indicate that the conventional methods are quite conservative relative to the more exacting time history analysis. They also indicate that the rocking motion is less on the foam crush pad than on the solid stainless steel pad

[en] In order to establish the capability of the TN24 cask for storage of spent fuel assemblies which are beyond the limits given by the manufacturer, a calculations of the dose and heat decay have been made for several cases of burnup higher than 35 GWd/MTU, using the SCALE 4.2 code package. The results were compared with the data obtained from the manufacturer. According to the results of the ORIGEN and SAS4 calculations and taking into the account limitations of the used model, it is possible to estimate that for 50 GWd/MTU burnup at least 15 years cooling time period is necessary to allow the use of TN24 cask. (author)

[en] The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source (60 Co radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

[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] Kobe Steel has been involved in the design, safety analysis and fabrication of transport and/or storage casks for radioactive materials for more than 20 years. Transport casks were primarily developed early on, however, now production has largely shifted to storage casks. To make the casks as safe as possible, without huge added expense, the advanced types of casks have been and will be developed and new materials such as high performance neutron shields and neutron absorbing materials are being increasingly developed and used. (author)

[en] The parameter effect analyses of a freestanding spent fuel storage cask are performed for an artificial time history seismic acceleration generated by the basis on the US NRC RG1.60 response acceleration spectrum. This paper focuses on the structural stability by seismic loads to check the overturing possibility of storage cask and the slipping displacement on concrete bed. Parametric analyses of a simplified cask model are performed to take into account the variations of seismic load in magnitude and interface friction between cask bottom and concrete bed. The analyses results show that the storage cask has a large marginal integrity in the response acceleration and slipping distance for various seismic loads and friction coefficients

No abstract available

[en] As initial enrichment and burnup of fuel assemblies in nuclear power reactors increases, taking burnup credit is now proceeding in LWR spent fuel for cask critically safety design. This paper describes a reactivity measurement system based upon the passive neutron measurement technique in wet condition (PA-w system) developed to obtain a subcriticality value by making direct in-cask measurement of neutron flux. The PA-w system comprises four sub-systems; the main control software, the computer analysis software, the data base and the on-site work. Analytical simulations, assuming nine PWR type fuel assemblies in a cask with a square array and four neutron detectors installed around the fuel assemblies, have been run to determine the system performance. The results demonstrated that the PA-w system is feasible for measuring the burnup of each loaded fuel assembly and the k-eff accurately, and has sufficient CPU time for fuel loading work. (authors)