Fuel pins, fast reactor

The Quantitative Description of Deformation and Stress in Cylindrical Fast Reactor Fuel Pins J. R. Matthews... 

THE QUANTITATIVE DESCRIPTION OF DEFORMATION AND STRESS IN CYLINDRICAL FAST REACTOR FUEL PINS... 

Summary of Computer Codes Being Applied to Liquid-Metal Cooled Fast Reactor Fuel Pin Thermal and Mechanical Behavior... 

A great variety of fast breeder fuel pins have been tested under operational transient conditions in sodium-cooled capsules in the High Flux Reactor (HFR) in Petten/Netherlands. The test field included power ramping, overpower, power-to-melt and temperature transient experiments and the irradiation was terminated end of 1993. In 1994 post-irradiation examinations and evaluation work were made leading to interesting results ... 

Critical Experiments with Fast Test Reactor Fuel Pins in Water, B. M. Durst, S. R. Bierman, E. D. Clayton (BNWL)... 

Criticality safety evaluations for handling mixed Pu-U oxide-type fuel elements depend heavily on computational analysis with experimentally validated computer codes and cross-section data. A series of critical experiments has been performed with fast test reactor fuel pins in water at the Battelle-Pacific Northwest Critical Mass Laboratory in support of the Advanced Fuel Recycle... 

Fig. I. Simplified description of fast test reactor fuel, pin. ...

Fig. 1. SimpUned dekription of the fast test reactor fuel pin.

During September 1978, four critical experiments involving fast test reactor fuel pins iimnetsed in water and partially reflected with concrete were performed at Battdle Pacific Northwest Laboratory (PNL). Documenitation of the experiments has been published elsewhere. Figure 1 shows the axial configuration of the pins. used in fire experiments. In all of the experiments, the pin arrays were surroimded with 0,40 m of concrete. Approximately 0.02 m of water was between the outehnost pins and the concrete reflector. 

Structural material decay heat. Heat is generated in the steels of the fuel pin cladding and the subassembly wrapper of a fast reactor fuel subassembly by the decay of radioactive products. This component of the decay heat does not exceed 10% of the total decay heat at any time. Integral measurements have been made of the activity induced in structural materials and these enable this component to be predicted to an accuracy of about 10%... 

I0.6.8.I Cladding failure in oxide fuel pins of nuclear reactors. The long-term operational performance of nuclear fuel pins is critically governed by the reactions that occur in the gap between the fuel and its cladding. Ball et al. (1989) examined this for the cases of (1) Zircaloy-clad pellets of U02+, in a pressurised water reactor (PWR) and (2) stainless-steel-clad pellets of (U, P)02+, in a liquid-metal-cooled fast-breeder reactor (LMFBR). In particular they were interested in the influence of O potential on Cs, I, Te and Mo and the effects of irradiation on the gaseous species within the fuel-clad gaps. 

Lead cooled fast reactor design assumes the subassembly with fuel pins arranged in square lattice with large pitch-to-diameter ratio (s/d 1.4-1.5). In LMFR a more tight arrangement of fuel pins is adopted (s/d 1.1-1.18) ... 

The experimental fast reactor JOYO at the Japan Nuclear Cycle Development Institute s Oarai Engineering Center attained initial criticality in April 1977 and was the first liquid metal cooled fast reactor in Japan. From 1983 to 2000, JOYO operated with the MK-II core as an irradiation test bed to develop the fuels and materials for future Japanese fast reactors. Thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. This paper provides a review of the operational experiences obtained through the JOYO s operation. 

Series of analyses were performed to investigate the basic characteristics of Pu burning in a fast reactor by changing various parameters including fuel pin specifications, smear density, core height, Pu vector, types of inert matrix without U, etc. 

On the in-pile experiments jointly conducted with French CEA, all the planned tests have completed for the CABRI-FAST program, in which slow and fast transient tests have been conducted, mainly with high bumup, annular fuel pins. A joint synthesis work is underway. Starting in 1996, a next joint in-pile test program, CABRI-RAFT, is initiated, where the total of 7 tests are planned in the CABRI and SCARABEE reactors through 2000. 

A Reference Design for the primary circuit of a 60 MWt experimental fast reactor is completed. Some parameters and characteristics have been selected mostly based on the Integral Fast Reactor (IFR) concept. Fuel pin dimensions and other data were used for calculations which lead to a general core configuration. Some calculations were also independently performed at lEN, and the differences in methodologies are being evaluated. Table 1 shows the main primary circuit parameters. 

Nonetheless, the experimental criticality data cur> rently available for LMFBR fuels do provide a secure validation point for (Pu, U)C>2 fuels to be used in the fast flux test facility (FFTF) program. Indeed, the data in Table H were obtained for this program. As a consequence, the criticality safety parameters for FFTF fuel pin handling are currently based directly on the experimental data, since the fuel rods used in the experiments are nearly the same sizes as those to be used in the FFTF reactor. These criticality parameters are as follows ... 

The Advanced Fuel Recycle Program is concerned with the safe reprocessing of mixed plutonium and uranium oxide fuels, characteristic of fast reactors. The safe handling and storage of these fuels hin on calculations and these in turn depend on clean, well-deflned experiment data for validation. Benchmark experiment data have been. acquired for fast test reactor (I R)-type fuels for impoisoned systems and systems intermixed with soluble poisons. However, there, are no data now available, fliat explore the criticality of these fuels intermixed with solid neutron absorbers (poisons). In this paper, we will present the results of experiments performed at the Pacific Northwest Laboratory (Critical Mass Laboratory) on fast test reactor fuel elements intermixed with solid neutron absorbers. The isons used were Bbral and cadmium plates and gadoliniuth cylindrical rods. Each absorber was separately examined to see its reactivity effect on lattices of FTR fuel pins in water none was intermixed. 

For metal fuel fabrication, the actinide metals are alloyed in an injection casting furnace that melts, mixes the alloy and injects the molten metal into quartz molds. After quick cooling, the quartz mold is removed from the metal pin, which is cut to length and undergoes quality assurance measurements. These pins are placed into new fuel cladding that contains a small amount of metallic sodium, which provides a thermal bond in early irradiation in the nuclear reactor. These fuel elements are welded closed and are ready for the reactor. Recent research in this area has focused on modifying the process to minimize the volatization of americium, which is a key component in U/TRU recovered for fast reactors and has a high vapor pressure. 

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. 

RAPID (Refuelling by All Pins, Integral Design) is the abbreviation for a small sodium cooled reactor of 10 000 kW(th) (1000 kW(e)) with U-Pu-Zr metal fuel and fast neutron spectrum [XVII-1 and XVII-2]. It is one of the successors of the RAPID-L [XVII-3 to XVII-7] - the operator-free fast reactor concept designed for a lunar based power system. 

The containment function is not vested in the reactor building or the earthen mound or the flibe heat transport pipes to the BOP. The first containment boundary is the fuel pin cladding. The second containment boundary is comprised of the reactor vessel, the refuelling cap on the reactor head, and the IHX tube walls. The third containment is the guard vessel and an extension of the vessel head - and (maybe) fast acting valves on the flibe loop pipes. 

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