Experimental Plan and Design of Two Experiments for Graphite Irradiation at Temperatures Up to 1500 Degrees C in the Target Region of the High Fluxisotope Reactor PDF Download
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Languages : en
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Two irradiation capsules have been designed for the target region of the high flux isotope reactor (HFIR). The objective is to provide dimensional change and physical property data for four candidate next generation nuclear plant (NGNP) graphites. The capsules will reach peak doses of ~1.59 and ~4.76 dpa, respectively, at temperatures of 900, 1200, and 1500 C.
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Languages : en
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Book Description
Two irradiation capsules have been designed for the target region of the high flux isotope reactor (HFIR). The objective is to provide dimensional change and physical property data for four candidate next generation nuclear plant (NGNP) graphites. The capsules will reach peak doses of ~1.59 and ~4.76 dpa, respectively, at temperatures of 900, 1200, and 1500 C.
Author: J. M. Davidson
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Category : Gamma rays
Languages : en
Pages : 104
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Category : Nuclear energy
Languages : en
Pages : 1004
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Languages : en
Pages :
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A high temperature graphite irradiation experinient was performed in the GETR core to determine the effects of differences in manufacturing, formulation, and graphitization temperatures on radiation-induced eontraction. The experiment was performed at temperatures of 800 to 1200 deg C in an intense fast neutron flux. The maximum integrated exposure of the sample positions was 3.2 x 10?sup 21/ nvt, E> 0.18 Mev, corresponding to approximately 24,000 MWD/AT in a conventional graphite-moderated reactor. All the graphites tested, with the exception of the controls, were needle coke filler, coal tar pitch binder graphites varying mn particle size, graphitization temperature, and impregnation. From theoretical and experitnental considerations, the formulations and treatments were expected to result in a relatively stable graphite in the direction transverse to extrusion. For comparison of the experimental results to existing experience, a conventional graphite, CSF, was used at each irradiation position. The results showed that the graphite most stable to contraction was graphaitized at a high temperature(>3100hC) and made from small particle size (all flour) filler. In all cases, the needle coke graphite contracted at a lower rate than the CSF graphite. Differences attributable to the size of extrusion and/or post graphitization cooling rate were discerned readily. Auxil iary to the purposes of the experiment, the apparent thermnal neutron cross section for Co/sup 58/ (plus Co /sup 58m) was determined. Co/sup 58/ and Co/sup 58m/ are the products of the Ni/sup 58/ (n,p) reaction, which is used widely for fast flux monitoring. Both have large thermal neutron capture cross sections which must be accounted for to prevent error in fast neutron dosimetry. In this experiment, a value was determined for the apparent burn-out cross section of 3750 barns. (auth).
Author: J. W. Helm
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Category : Graphite
Languages : en
Pages : 122
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Author: R. Manzel
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Category :
Languages : en
Pages : 17
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Category :
Languages : en
Pages :
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The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.
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Languages : en
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The irradiation behavior of reactor graphites at high temperatures (500° to 1400°C) to high fluences (up to 4 x 1022 n/cm2) is reviewed. Recent data generated during the period 1965 through 1971 are emphasized. The review covers graphite manufacture, irradiation damage, dimensional and structural changes, thermal expansion, thermal conductivity, Young's modulus, strength, and irradiation-induced creep.
Author: J. M. Davidson
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Category : Graphite
Languages : en
Pages : 50
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Languages : en
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Graphite specimens fully clad with Type 310 stainless, Type 446 stainless, or Inconel were irradiated in a 1000-psi CO/sub 2/ environment for a period of 4 weeks at approximately 1300 F followed by 1 week at 1500 F. The fast-neutron-plus-gmma dose rate was estimated at 8 x 10/sup 7/ rads per hr. The gas environment was sampled and replenished eight times during the experiment. After 60 hr at about 1350 F, the CO content had in creased from a negligible value to 3.6 vol. %. It then decreased steadily to a value of 0.4 vol. % at the end of 4 weeks. When the temperature was raised to 1500 F, the CO content increased to 1.9 vol, % in 100 hr and then decreased to 0.6 vol. %. The O content remained nearly constant throughout the experiment. Postirradiation examination and metallography revealed very little effect of the exposure on the cladding materials at the gas and graphite interfaces. (auth).
