Fuel irradiated, swelling

The Rapsodie experimental sodium cooled reactor was the first French fast neutron reactor. The construction was started in 1962 within an association of CEA and EURATOM. The reactor went critical on 28 January 1967, reaching 20 MW (th) power on 17 March 1967. The core and equipment were modified in 1970 to increase the thermal power level to 40 MW (th). The operating parameters were similar to those in large commercial size reactors. During 16 yets of operation 30 000 fill pins of the driver core were irradiated, of which -10 000 reached a bumup beyond 10% 300 irradiation experiments and more than 1 000 tests have been performed. The maximum bumup of the test fuel pins was 27% (173 displacement per atom). In 1971, the irradiations performed in the core revealed a phenomenon of irradiation swelling in the stainless steel of the wrapper and the fuel cladding in the high neutron flux. The R sodie results have been extrapolated in the Phenix reactor. 

A.S22. The assessment shall take into account changes to safety relevant fuel parameters that may be caused by mechanical deformation, irradiation swelling, etc., as mentioned in paras A.506 and A.508. 

In fire presence of a mechanical load, for example the fission gas pressure in a fuel pin or the sodium pressure inside a hexagonal can, another irradiation-induced type of deformation can occur the so-called irradiation creep phenomenon that is a macroscopic plastic strain occurring along the direction of the stress. Irradiation swelling and creep are correlated because both of these phenomena are exclusively related to the evolution toward the point defects of the sinks of the vacancies and interstitials generated in the atomic colhsion cascades produced inside the crystal lattice by the... 

The amount of plutonium that can be fissioned before the fuel has to be removed from the reactor for either loss of reactivity, fuel element swelling, or irradiation damage that threatens reactor safety must be considered. Weapons-grade plutonium generates approximately one noble gas atom for every four fissions. At very high bumups, this will cause the fuel to swell to volumes that are unacceptable for most traditional fuel forms. 

A considerable amount of operating experience with U-Al fuels in aluminum and aluminum alloy cladding exists. This should be at least partially applicable to the Pu-Al fuels. The maximum bumup level expected with Pu-Al fuels without a recycle step is based on the performance of U-Al fuel plates irradiated in the Advanced Test Reactor. This bumup level (70% FIMA) is based on fuel element swelling and fission product release limitations in a thin plate geometry under conditions of high fission rates. More robust geometries might be fissioned to safe bumup levels in excess of 90%. Burnable poisons have been added to U-Al fuel plates with little difficulty, so burnable poisons could probably be added to a Pu-Al fuel form. 

Fast breeder reactor fuel rods consist of stainless-steel-clad mixed oxide (U,Pu)02 fuel however, more stable alloys for cladding and in-core structural materials, with resistance to swelling and embrittlement under fast neutron irradiation, and more efficient fuels (carbide see 17.3.12.1.2) or nitride (see 17.3.12.3)] are needed h The mechanical, metallurgical, and chemical processes in fuel element irradiation are depicted in Figure 1. Figure 2 shows the PFR (U.K.) fast breeder fuel element, and Figures 3 and 4 illustrate the Fast Flux Test Facility (FFTF) fuel system. 

The choice of cladding material for fast reactors is less dependent upon the neutron absorption cross section than for thermal reactors. The essential requirements for these materials are high melting point, retention of satisfactory physical and mechanical properties, a low swelling rate when irradiated by large fluences of fast neutrons, and good corrosion resistance, especially to molten sodium. At present, stainless steel is the preferred fuel cladding material for sodium-cooled fast breeder reactors (LMFBRs). For such reactors, the capture cross section is not as important as for thermal neutron reactors. 

Expansion of the fuel pellets due to high internal temperatures, cracking due to thermal stresses, and irradiation-induced swelling may lead to contact of the fuel with the cladding. 

Densification, which is the reverse of swelling, is a result of irradiation. Such behavior can cause the fuel material to contract and lead to irregularities in the thermal power generation. 

In a specific fuel pin irradiation (KNK ll-VENKER) it could be shown that the temperature gradient in the cladding will increase swelling by as much as 50 % compared to material irradiations this is a consequence of he He-bubbles migration in the temperature gradient and leading to an accelerated void... 

Reactor core Basic difficulties in achievement of reliable and effective operation of the BN-350 core were connected with some irradiation impacts on stmcutural material properties, selection of non-optimal materials, which under the irradiation of about 50-60 displacement per atom (dpa) suffer a considerable swelling and fragility that cannot allow continued normal operation of hexagonal tubes of sub-assemblies, cover tubes of absorbing assemblies and their elements. The results of a special research programme on irradiated materials allowed consequently to improve the BN-350 core and increase maximum fuel bum-up from 5% of heavy atoms up to 7%, then 8.3% and now about 12%. Simultaneously, the life time of absorbing assemblies was increased from 60 to 450 effective days. 

The choice of mononitride fuel for fast reactors was dictated by its certain, if small, advantages over carbides from the viewpoint of density, swelling, and confinement of fission gas. An important factor in the final decision made in favor of mononitrides was the higher pyrophorosity of monocarbide, which causes problems in fuel fabrication and handling of irradiated fuel with failed claddings (Rogozkin et al. 2003). 

The plant is designed for on-load refuelling. Increasing the aluminium concentration in the uranium fuel rod to reduce the rate of swelling has permitted longer irradiation. An average discharge burnup of 5300 MWd/t without any limit on residence time in the reactor core is now achieved. 

Furthermore many studies and improvements on fuel and cladding to limit risks of blockings have been performed to limit buckling (mechanical axial clearance, severe restrictions on moisture in U02 pellets, etc.) and irradiation induced swelling (increasing of oxide densification, etc.). 

Analyses shall be performed to show that the intended irradiation conditions and limits (such as fission density, total fissions at the end of lifetime and neutron fluence) are acceptable and will not lead to undue deformation or swelling of the fuel elements. The anticipated upper limit of possible deformation shall be evaluated. These analyses shall be supported by data from experiments and from experience with irradiation. Consideration should be given in the design of the fuel elements to the requirements relating to the long term management of irradiated elements. 

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