Spent fuel management

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International Approaches to Spent Fuel Management: Challenges and Opportunities

International Approaches to Spent Fuel Management: Challenges and Opportunities

Spent fuel management in many European nations is considered a ‘European challenge.’ This is particularly true of those with relatively small nuclear power programs although several countries with advanced repository programs have expressed some concern that any focus on regional solutions could undermine domestic support for their own national repository programs. But for the majority, this has nurtured collaboration. From 2003 to 2008, the European Commission funded pilot studies on the feasibility of shared regional storage facilities and geological repositories in Europe and options for the establishment of a European Repository Development Organization (ERDO). A working group – which has included representatives from Austria, Bulgaria, Denmark, Ireland, Italy, Lithuania, Netherlands, Poland, Romania, Slovakia and Slovenia since its creation in 2009 – has been established to build consensus about how to proceed with cooperative approaches to spent fuel management and lay the ground work for the establishment of the ERDO. While the sensitive issue of siting has been deliberately postponed until more work has been completed on both geological screening and trust-building, the working group serves as an important forum to exchange information and build national capacity.[10] On July 19, 2011 the European Council adopted a legally binding and enforceable European Commission proposal which read in part: “Radioactive waste shall be disposed of in the Member State in which it was generated, unless agreements are concluded between Member States to use disposal facilities in one of them.”[11] It should be noted that the complexities of managing liability and other legal arrangements are minimized to a significant extent in the EU and Euratom due to an existing common understanding of basic concepts and an agreed regulatory framework, making the application of the EC example elsewhere complicated.
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Sixth situation report on radioactive waste and spent fuel management in the European Union. Commission staff working document accompanying the report. SEC (2008) 2416 final/2, 16 July 2010

Sixth situation report on radioactive waste and spent fuel management in the European Union. Commission staff working document accompanying the report. SEC (2008) 2416 final/2, 16 July 2010

The first choice facing Member States is their choice of spent fuel management policy i.e. reprocessing or direct disposal. The first option will recover plutonium and uranium for possible re-use, but also generate HLW, LILW-LL and LILW-SL, all of which will require disposal. In the case of the first two categories, this should take place in a deep geological repository. Currently five states make use of the reprocessing option; Bulgaria, France, Germany (in the case of the remaining spent fuel at reprocessing facilities), the Netherlands and UK. Italy also intends to reprocess the remaining fuel from its closed reactors. If current plans are pursued, Germany will no longer reprocess fuel once the current contracts expire. The UK is still keeping open the option of new business for Thorp, but any new contracts would need Government approval. Belgium has a moratorium on new reprocessing contracts since 1993. In the past Spain exported a small amount of fuel for reprocessing, but has since stored all fuel at its NPPs (Spain is planning a centralised storage facility for HLW and spent fuel to be operational before 2011).
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Comparative Analysis of VVER 1000 Westinghouse and TVEL Spent Fuel Capability

Comparative Analysis of VVER 1000 Westinghouse and TVEL Spent Fuel Capability

It is necessary to note slightly more extensive heat release in FA-WR (Westinghouse) compared to TVS-A (TVEL) under the same burnup depth. This can cause the need for slightly longer post-operation fuel cooling in spent fuel pool. In general, obtained results allow making conclusion that from the viewpoint of safe spent fuel management and storage implementing new alternative fuel of Westinghouse company at VVER-1000 does not require modifications of current conditions and procedures. For the majority of considered spent fuel characteristics the differences between TVS-A (TVEL) and FA-WR (Westinghouse) are less than total changes of these characteristics depending on production tolerance and operation conditions.
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Rural Livelihoods, Forest Access and Time Use: A Study of Forest Communities in Northwest India

Rural Livelihoods, Forest Access and Time Use: A Study of Forest Communities in Northwest India

According to the Forest Survey of India, forests covered 42.35 percent of the total geographical area of Mandi district (FSI, 2009). Following the British legacy of forest nationalisation, the state is responsible for managing forests and wildlife. Unlike in other parts of India, the settlement of forest rights in Mandi district (which began after 1878 and ended in 1917) did not lead to a full termination of traditional forests rights allowing all households defined rights to their community forests (Saberwal, 1999; Chhatre, 2003; Vasan, 2003; Chhatre and Saberwal, 2006). Nevertheless, the exercise of these rights is filtered through the legitimacy allowed by the state. Protected areas, including wildlife sanctuaries and national parks, have the most restricted access. The study region consists of a wildlife sanctuary that continues to allow human settlements. Residents of the sanctuary extract forest products, even though the Forest and Wildlife Departments monitor the sanctuary and impose sanctions on violators. This policy increases the costs of access to forest products and community management relative to residents in
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MICROBIAL FUEL CELL (MFC) IN TREATING SPENT CAUSTIC WASTEWATER

MICROBIAL FUEL CELL (MFC) IN TREATING SPENT CAUSTIC WASTEWATER

From Table 1 and Table 2, it is observed that spent caustic wastewater is containing high concentration of hazardous contaminants. Due to the presence of these contaminants, the exposure of spent caustic wastewater is harmful to both human and environment. The possible exposure route for spent caustic wastewater could be through eyes, ingestion and inhalation. According to its Material Safety Data Sheet. (2012), exposure via the eyes would cause irritation to the eyes and eye burns. Dermally contacted with spent caustic would cause irritation and burns to the skin which can be characterized by itching, scaling, reddening or blistering. The ingestion of spent caustic would cause burns to oral cavity, lips, upper airway, oesophagus and possibly the digestive tract. Spent caustic solution would cause severe irritation to the respiratory system and may cause burns to the respiratory tract and mucous membranes (MSDS, 2012). Since spent caustic wastewater is highly toxic, the discharge of spent caustic to the environment would impact the biological system. For example, the release of spent caustic into water bodies would cause the pH, TSS and COD concentration to increase and that would create a no longer suitable condition for the survival of the aquatic life. The discharge of raw spent caustic waste will not only affects the water bodies, but it will also likely to be volatile in the air thus affecting the air quality as well. The livings might end up to breath in the harmful air which would cause them to suffocate and could not survive the condition. Also, spent caustic wastewater might as well remains and sediment in the soils. This will impact the living organisms in the soil and plant. Plants might not be able to grow well as it absorbs harmful substances from the soil. Not only that, the consumers of the plant could have consumed poisonous plant and cause their health to be severely impacted. Therefore, to ensure human and environment not to be effected by spent caustic wastewater, safety handling and disposal of spent caustic is very crucial.
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Overview of Independent Spent Fuel Storage Installations (ISFSIs) in the New Madrid Seismic Zone (NMSZ) and Wabash Valley Seismic Zone (WVSZ)

Overview of Independent Spent Fuel Storage Installations (ISFSIs) in the New Madrid Seismic Zone (NMSZ) and Wabash Valley Seismic Zone (WVSZ)

Facility, NUREG-1536, and the Standard Review Plan for Spent Fuel Dry Storage Facility, NUREG-1567 provide the NRC staff guidance in reviewing the license applications, and suggest that the storage cask be designed not to tip-over for all man-made events and external natural phenomena. However, consistent with the NRC’s defense-in-depth policy, these guidelines suggest that the DCSS structural integrity should be evaluated for a non-mechanistic tip-over event. An independent structural analysis of a DCSS was performed by the staff, for a cask tip-over event using the explicit method of dynamic analyses in the finite-element (FE) computer code LS-DYNA, Version 960. The analysis was performed for a cask angular velocity of 1.7 radians per second, at the time of impact on the pad (see Figure 3 and Figure 4) The cask angular velocity of 1.7 radians per second was based on the gravity fall with an initial zero velocity at the start of the tip-over (center of gravity over the cask corner), plus an additional 10 percent increase to account for a potential for a non-zero initial velocity during an
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Localization of Plastic Deformation in Copper Canisters for Spent Nuclear Fuel

Localization of Plastic Deformation in Copper Canisters for Spent Nuclear Fuel

Localization of plastic deformation in different parts (extruded and forged base materials as well as EB and FSW welds) of the corrosion barrier copper canister for final disposal of spent nuclear fuel was studied using tensile testing, optical strain measurement, scanning electron microscopy (SEM), and electron back-scatter diffraction (EBSD). Results show that in the base materials plastic deformation occurs very uniformly. In FSW welds the deformation localizes in the weld either at the processing line or at a line of entrapped oxide particles. In EB welds the deformation localizes to the equally oriented large grains at the weld centreline or at the steep grain size gradient in the fusion line.
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Structural Integrity Evaluation of Spent Fuel Storage Racks for HANARO

Structural Integrity Evaluation of Spent Fuel Storage Racks for HANARO

To evaluate the fuel assembly rattling phenomenon, the fuel assembly is modeled using a beam element, and the gaps between fuel assembly and cell pipe are modeled using a 3-dimensional combination element in ANSYS. It is assumed conservatively that the gap between the fuel assembly and the cell pipe changes from a maximum of twice the nominal gap to a zero gap. The fuel assembly rattling analyses are performed for four analysis cases using the displacement time histories of the cell pipe bottom end in modules obtained from the seismic analysis. Only the displacement time history for the case of the top module is selected for an analysis because it seems to be more severe than those for other module cases. These displacement time histories are applied at the fuel bottom and cell pipe top, simultaneously.
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Stability Analysis of Storage of Spent fuel in Stack of Trays in Pool

Stability Analysis of Storage of Spent fuel in Stack of Trays in Pool

A finite element model of stack of spent fuel trays has been developed to represent the behaviour under seismic condition. The models are created for each individual tray and spent fuel bundles are modelled as a lumped mass in tray FE model. The interlock between trays has been simulated using combination (i.e. spring + Gap) element. The contact element has been used in FE model to capture the lifting and sliding motion between trays and the same has been used for contact between resting surface and bottom most tray of stack. A time history analysis was performed for stack of 20, 25, and 30 trays with different friction of coefficient (0.1, 0.2 & 0.3) to verify its overall stability against turning and sliding under seismic event. These stacks of trays are found stable under the designed level of earthquake.
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Analysis of Raft Foundations for Spent Fuel Pool in Nuclear Facilities

Analysis of Raft Foundations for Spent Fuel Pool in Nuclear Facilities

Foundation rafts are analysed as a plate on elastic foundation with the representation of the foundation media using the Winkler idealisation i.e. series of linear uncoupled springs. The elastic constant of the Winkler springs is derived using the sub-grade modulus. However, the Winkler approach has limitations due to incompatibility of the deflections at raft-soil interface. The deflection of the raft at the point of contact and the deformation of the foundation media at this point of contact are incompatible in this approach. This particularly influences flexible rafts and further if the foundation media is soil. This paper discusses the analysis of raft, in general, and the analysis of the foundation raft for a Spent Fuel pool facility using ‘variable k approach’ where deformations at a node and influencing nodes are computed using Boussinesq’s theory. The limitations stated above are overcome in this approach. Some studies on the sensitivity of parameters were carried out in the form of variation of modulii of elasticity of concrete and deformation modulus of soil. Analysis is also performed with conventional method using ‘Winkler’ soil springs.
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Risk assessment of spent fuel ponds subject to the risk of flooding

Risk assessment of spent fuel ponds subject to the risk of flooding

Built between 1988 and 1995, the power plant includes two main turbine generators and a single reactor based on a Westinghouse standard four loop pressurised water design. The initial design was modified, mainly in terms of capacity and redundancy of safety system, in order to fulfil UK requirements. The on- site electric substation is connected to the external grid at three separate 400kV points (two at Bramford, one at Norwich and one at Pelham) and provides connection with the external network for the import and export of power. Adjacent to the reactor building, the fuel building accommodates the pond where both new and used fuel is stored [Fullalove (1995)] under water. The pool consists of a stainless steel lined reinforced cavity where the fuel assemblies are located at a depth of water adequate to guarantee the coverage of the fuel for 24h in case of total loss of the cooling system. The latter consists of a primary ultimate heat-sink (seawater) and a reserve ultimate heat-sink (air-cooling system) which ensure the thermal exchange required for the pumped flow. The availability of AC power on-site binds the working order of the cooling system in the fuel facilities. These, as all the building of the nuclear island, are provided with fire doors that can act as flood barriers up to a water depth of 1m [EDF (2012)].
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Buckling Behavior of Spent Nuclear Fuel Rods under Impact Loading

Buckling Behavior of Spent Nuclear Fuel Rods under Impact Loading

Previous studies have addressed fuel rods buckling capacity, but including simplifications for PCI in their “g” load calculation. For instance, Sanders et al. (1992) used only the percentage of fuel pellet mass effective in loading the cladding. They considered that roughly 75% of the fuel pellet mass was attached to the cladding in pressurized water reactors (PWRs). For boiling water reactors (BWRs) the effective pellet mass attached to the cladding was assumed as 10%. These assumptions are based on evidence of BWR SNFs showing minimal PCI, whereas in PWR model fuel pellets are sufficiently attached to the cladding. Based on the assumed percentage of fuel pellet mass attached to cladding, the number of “g’s” required to cause buckling of the cladding was calculated. Ramsey et al. (1987) only considered the weight of cladding and neglected the weight of fuel pellets in their analytical calculation for “g” load necessary for axial buckling. In addition, Bjorkman (2010) concluded that the impact “g” load magnitude of a fuel rod under inertia loading depends on the weight of the cladding and fuel rod, regardless of whether the fuel is bonded to the cladding.
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Seismic Stability Analysis for an In-Bay Worktable for Spent  Fuel Dry Storage

Seismic Stability Analysis for an In-Bay Worktable for Spent Fuel Dry Storage

The in-bay worktables are part of the spent fuel dry storage system. Figure 1 shows a sketch of the in-bay worktables. The function of the worktables is to provide a movable work surface and structural support to the equipment used in the basket loading operations (i.e., the fuel bundle tilter, the turntable, and the transfer table). It is necessary that the free standing underwater worktables maintain its structural integrity and stability under a severe earthquake. The objective of the analysis is to seismically qualify the stability and integrity of these free standing worktables under the Design Basis Earthquake (DBE). Since the worktables are submerged in water, hydrodynamic effects must be considered in the analysis to more accurately predict their structural behaviours. In addition, as the worktables are free standing type structures, care must be taken in the development of the finite element model and in the interpretation of the analysis results.
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Seismic   Analysis and Evaluation of Spent Fuel Dry Cask Storage Systems

Seismic Analysis and Evaluation of Spent Fuel Dry Cask Storage Systems

The soil profile data indicate that this site has a soft soil foundation° Since the seismic input amplification was found to be rather sensitive to the choice of [r]

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Seismic Analysis of Free-Standing Dry Storage Cask for Spent Fuel

Seismic Analysis of Free-Standing Dry Storage Cask for Spent Fuel

For most dry storage facilities for spent fuel, storage casks are freestanding on a concrete pad. Thus, relative motion between the cask and the pad may be induced during earthquakes, leading to stability concerns. In this study, the seismic stability conditions of a freestanding cask were examined quasi-statically, and the borderline value friction coefficient which differentiates the dominant motion type between sliding and rocking was obtained. Then, a finite element (FE) cask-pad model considering the frictional contact at their interface was established, and a dynamic analysis simulating a scaled cask shaking table test was conducted. Through appropriate settings, the nonlinear dynamic analysis can reasonably reproduce the test results. The influence of the friction coefficient of the cask/pad interface on the cask response was further investigated using this FE model. For the case with lower friction coefficient, the sliding response dominated the seismic motion of the cask; while apparent rocking of the cask was caused at a higher friction coefficient. Thus, the results of the FE analysis are conformable to the quasi-static analysis. In addition, the nutation motion of the vertical cylindrical cask (VCC) which follows the rocking response was also discussed, which may possibly cause considerable horizontal displacement and is therefore unfavorable to the seismic stability of the VCC.
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Seismic Response Analysis of a Freestanding Model of Spent Fuel Storage Cask

Seismic Response Analysis of a Freestanding Model of Spent Fuel Storage Cask

The seismic response analyses of a freestanding spent fuel storage cask are performed for an artificial time history acceleration generated on the basis of the US NRC RG1.60 response acceleration spectrum. This paper focuses on the structural stability regarding seismic loads to check the overturning possibility of a storage cask and the slip displacement on the concrete installation bed. A simple structural analysis model for the storage cask is developed to perform the parametric effect analyses regarding the seismic responses. Two parameters considered in the analyses are the magnitude of the seismic load and the interface friction between the cask’s bottom surface and the upper surface of the concrete installation bed. The analyses results show that the seismic responses of the storage cask are influenced by a combination of the two parameters and the storage cask also has a large marginal integrity for the maximum overturning angle and the slip distance for the design and beyond design seismic loads. Keywords: Spent fuel storage cask, seismic response analysis, lumped-mass model.
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Proposal Sites for Spent Fuel Disposal in Egypt by Using GIS Program

Proposal Sites for Spent Fuel Disposal in Egypt by Using GIS Program

ABSTRACT: The present study uses Geographic Information System(s) (GIS) as a spatial decision support tool to select potential sites in Egypt for deep geological disposal for spent radioactive fuel. A specific weight is given to each criterion according to its relative influence on the process of decision making. The results from the application of the presented methodology determine the relative suitability of sites for deep disposal of radioactive wastes. These sites are ranked in descending order to help decision-makers in Egypt. The present study aims to provide a baseline for more detailed technical analyses for the selected sites.
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A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel

A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel

A unique feature of nuclear energy is the availability in spent nuclear fuel of recyclable fissile and fertile materials able to provide new fuel to generate power. When we recycle paper, glass, plastic or metal we separate useful materials from waste, mainly to reduce the consumption of fresh raw materials and also to diminish the air and water pollution. Spent nuclear fuel contains considerable amounts of useful materials. To illustrate this, the composition of used nuclear fuel is presented in the first section of this chapter. Furthermore, an overview of reprocessing is given to bring this thesis into context. A comparison of reprocessing methods, underlining the advantages and disadvantages of current electrochemical processing technology, is provided for a better understanding of the motivation of using and improving this technology.
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EFFECT OF THE FLUID COUPLING BETWEEN THE SPENT FUEL STORAGE RACK AND THE FLOOR

EFFECT OF THE FLUID COUPLING BETWEEN THE SPENT FUEL STORAGE RACK AND THE FLOOR

In general, the spent fuel rack structure consists of a matrix of multi cells. These cells are welded to a baseplate, providing horizontal and vertical support to the cells. Since the spent fuel rack is free standing, the rack support legs can rest, slide, tilt, or lift off depending on the inertia of the rack and the friction on the floor during base excitation. This subject has been studied by several researchers [Ren and Stabel (1999), Scavuzzo et al. (1979), Fritz (1972), and Zhao et al. (1996)], and the effective approach has been established by using a hydrodynamic mass matrix to account for the fluid-structure coupling. For the current level of earthquakes, the horizontal movement of the rack is higher than the vertical movement; therefore, only the horizontal fluid-structure coupling is considered (based on the fact that the vertical motion is relatively small or zero due to the consideration of gravity). However, as seen in the 2011 Fukushima earthquake in Japan, a future earthquake may be higher than the design basis earthquake, causing the rack to tilt or lift off. Thus, in this paper, an exploratory study regarding the effect of the vertical hydrodynamic coupling is performed. In general, prior to lifting off or rocking, rack motion is dominated by sliding. However, once the rack lifts off or tilts, the motion of the rack is governed by twisting rather than sliding, which leads to highly nonlinear dynamic behavior of the rack. To keep the focus on the effect of the vertical hydrodynamic coupling, rack motion is limited to rest, rocking, sliding, or lifting off by using a two-dimensional beam model that prevents the rack from twisting.
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Tip-Over Analysis of the   HI-STORM Dry Storage Cask System (B237)

Tip-Over Analysis of the HI-STORM Dry Storage Cask System (B237)

The HI-STORM 100 dry-cask storage system, shown in Figure 1, consists of a sealed metallic canister, referred to as the Multi-Purpose Canister (MPC), contained within a steel-concrete-steel overpack. The HI-STORM 100 System is designed to accommodate a wide variety of spent nuclear fuel assemblies in a single overpack design by utilizing different MPCs. While the external dimensions of all MPCs are identical to allow the use of a single overpack, each of the MPCs has different internal baskets to accommodate distinct fuel characteristics. Each MPC is identified by the maximum quantity of fuel assemblies it is capable of storing. The MPC-24 contains a maximum of 24 PWR fuel assemblies, and the MPC-68 contains a maximum of 68 BWR fuel assemblies.
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