Reactor fuel bundle

Uranium oxide [1344-57-6] from mills is converted into uranium hexafluoride [7783-81-5] FJF, for use in gaseous diffusion isotope separation plants (see Diffusion separation methods). The wastes from these operations are only slightly radioactive. Both uranium-235 and uranium-238 have long half-Hves, 7.08 x 10 and 4.46 x 10 yr, respectively. Uranium enriched to around 3 wt % is shipped to a reactor fuel fabrication plant (see Nuclear REACTORS, NUCLEAR FUEL reserves). There conversion to uranium dioxide is foUowed by peUet formation, sintering, and placement in tubes to form fuel rods. The rods are put in bundles to form fuel assembHes. Despite active recycling (qv), some low activity wastes are produced. 

Tong, L. S., 1967b, Heat Transfer in Water-Cooled Nuclear Reactors, Nuclear Eng. Design (5 301. (3) Tong, L. S., 1968a, An Evaluation of the Departure from Nucleate Boiling in Bundles of Reactor Fuel Rods, Nuclear Sci. Eng. 33 7-15. (5)... 

Enriched UF is shipped to the plant for fabricating reactor fuel elements in monel cylinders whose size is determined from the content, so as to prevent accumulation of a critical mass. At the fuel fabrication plant UF is converted to UO or other chemical form used in reactor fuel. For light-water reactors the UOj is pressed into pellets, which are sintered, ground to size, and loaded into zircaloy tubing, which is filled with helium and closed with welded zircaloy end plugs. These individual fuel rods are assembled into bundles, constituting the fuel elements shipped to the reactor. Conversion of UFj to UO2 is described in Chap. 5. Extraction of zirconium from its ores and separation of zirconium from its companion element hafnium is described in Chap. 7. 

D. C. Groeneveld and W. W. Yousef, Spacing Devices for Nuclear Fuel Bundles A Survey of Their Effect on CHF, Post CHF Heat Transfer and Pressure Drop, Proc. ANSIASMEINRC Information Topical Meeting on Nuclear Reactor Themtal-Hydraulics, Nuclear Regulatory Commission/CP-0014 (2) 1111-1130,1980. 

L. S. Tong, An Evaluation of the Departure From Nucleate Boiling in Bundles of Reactor Fuel Rods, Nuclear Sci. Eng. (33) 7-15,1968. 

The application of the mixed uranium-plutonium fuel in power reactors requires assurance of safe transport of semifinished items, fuel elements, and fuel bundles (FB). To research various aspects of safety, it is necessary to take into account that the thermal and radiation characteristics and criticality parameters of MOX fuel are higher than the characteristics of fuel on a basis of uranium dioxide. 

Besides, surfaces of FB and the fuel elements can be polluted by plutonium. The level of heat release for FB made of weapon plutonium is essentially higher than that for FB made from uranium (in the latter, heat release is practically absent). For example, heat release of FB of reactor BN-600 reaches 20 watts and that in FB of reactor VVER-1000 reaches 130 watts. The heat release of fuel bundles with regenerated or recycled reactor-grade plutonium is several times higher. 

Spent fuel bundles with fuel from power (i.e., reactor-grade) plutonium, which were stored for 1-3 years, are characterized by a heat release approximately 70% higher than those with uranium dioxides. The heat release from fuel bundles made of weapon plutonium is approximately 30% higher than that of FB with uranium [1]. 

VVER-1000 REACTOR FUEL ELEMENTS AND FUEL BUNDLES... 

At the moment, there are no transport packages in Russia that are suitable for the transportation of fuel elements and fuel bundles of reactor VVER-1000 with fresh mixed fuel. VNIPIET has performed design studies for such packages and appropriate auxiliaries. The work was conducted in two directions ... 

The safety of transport of MOX fuel requires careful analysis and substantiation. A use of available transport packages for MOX fuel transport will require essential upgrades and the development of new packages. This work would require considerable financial expense and a long time. Therefore, it should be started right now, so that there would not be a delay in deliveries of sample or regular fuel bundles to nuclear power plants. Besides, variants of the transport-technological methods of operation with fresh and spent MOX fuel in NPPs with reactor types VVER-1000 and BN-600 should be worked out. 

The radioactive products contained in the fuel are normally located in the sinterized uranium dioxide of the reactor fuel (the uranium dioxide fuel is shaped into pellets, roughly 1 cm in diameter, inserted in long zirconium alloy (zircalloy) cylinders). The matrix of these cylinders (roughly 40000), grouped in bundles to form the fuel elements, is the reactor core. 

The general aim of the Mol 7C e3q>eriments in the BR2 reactor was to investigate the failure events and the inherent coolability conditions in a fuel bundle subjected to a local (partial) coolant flow blockage in the fissile zone. Operation of reactors with small number of failed fuel rods is desirable for economic reasons, but safety must nevei be prejudiced. Therefore the behaviour of failed fuel during accident is of primary interest and fuel simulation codes are essential in establishing an operational set of safety measures. The present MOL7C experiments fit clearly the current need for improved techniques which should help establish operational limits for fuel rods and develop higher performance materials. 

To maintain efficient performance over the reactor volume for nominally 1,000 MWe LWR, about one third or about 251 of spent fuel is removed from the reactor core every year or so, the spent fuel being replaced with fresh fuel. However, over time, the concentration of fission fragments and heavy elements in a fuel bundle will increase to the point where it is no longer practical to continue to use the fuel. Therefore, after 24—36 months the core is removed from the reactor. 

In the event that a defect occurs in a fuel bundle during reactor operation, the fuelling machines can be used to remove the defective fuel, thereby limiting the release of fission products to the heat transport system coolant. Systems are provided for the detection and location of defective fuel. 

The CANDU 3 reactor includes 232 horizontal fuel channels. Each fijel channel consists of a zirconium-niobium alloy pressure tube, centred in a Zircaloy calandria tube by annular spacers, and expanded into a stainless steel end fitting at both ends. Each channel contains twelve fuel bundles. 

On-power fuelling is a feature of all PHWRs which have veiy low excess reactivity. In this type of reactor, refuelling to compensate for fuel depletion and for overall flux shaping to give optimum power distribution, is carried out with the help of two fuelling machines, which work in unison on the opposite ends of a channel. One of the machines is used to fuel the channel while the other one accepts the spent bundles. In addition, the fuelling machines facilitate on-power removal of failed fuel bundles. 

Maintain the neutron flux profile in the reactor close to its design shape so as to enable operation at the minimum possible power without exceeding the limits on fuel bundle power (flux tilt control) ... 

The. design of the Zero Power Test core requhed. at the system first be. built up with dummy fuel bundles. Multiplication runs were pafbtm by replacing fhese bundles with hiel bundles. Hie system was brought critical bsised on the reciprocal multipUcation curve and achieved criticality with 700 fuel elements a critical mass of 7.9 kilograms udtfa no Martin Power Reactor tods in the core. 

The fuel, assemblies studied were composed of 28-pin fuel bundles mounted in two concentric tubes simulating the pressure and caiandria tubes of the power reactor. Each fuel pin consisted qf a 47.7-cm len of sintered UPi pellets (1.42-cm diam) of density 10.45 g/cm sheathed in Zircaloy-2 (l.S2-cm OD X 0.45-mm wall). Symmetric rings of 4, 8 and 16 pins on diameters of 2.33, 5.30 and 8.41 cm respectively formed the 28-pin bundle. Aluminum mounting plates and Zircaloy end caps increased the over-all bundle length to 49.7 cm. Five bundles mounted in a 6SS-A1 pressure tube (12.78-cm OD X 2.96-mm wall) and a 50S-A1 caiandria tube (12.74-cm OD X 1.39-mm wall) formed a fuel assembly. 

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