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OSMOSE experiment representativity studies.

Description: The OSMOSE program aims at improving the neutronic predictions of advanced nuclear fuels through measurements in the MINERVE facility at the CEA-Cadarache (France) on samples containing the following separated actinides: Th-232, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, Cm-244 and Cm-245. The goal of the experimental measurements is to produce a database of reactivity-worth measurements in different neutron spectra for the separated heavy nuclides. This database can then be used as a benchmark for integral reactivity-worth measurements to verify and validate reactor analysis codes and integral cross-section values for the isotopes tested. In particular, the OSMOSE experimental program will produce very accurate sample reactivity-worth measurements for a series of actinides in various spectra, from very thermalized to very fast. The objective of the analytical program is to make use of the experimental data to establish deficiencies in the basic nuclear data libraries, identify their origins, and provide guidelines for nuclear data improvements in coordination with international programs. To achieve the proposed goals, seven different neutron spectra can be created in the MINERVE facility: UO2 dissolved in water (representative of over-moderated LWR systems), UO2 matrix in water (representative of LWRs), a mixed oxide fuel matrix, two thermal spectra containing large epithermal components (representative of under-moderated reactors), a moderated fast spectrum (representative of fast reactors which have some slowing down in moderators such as lead-bismuth or sodium), and a very hard spectrum (representative of fast reactors with little moderation from reactor coolant). The different spectra are achieved by changing the experimental lattice within the MINERVE reactor. The experimental lattice is the replaceable central part of MINERVE, which establishes the spectrum at the sample location. This configuration leads to a uniform well-behaved system so that the reactor configuration is in the fundamental mode. In fact, an important ...
Date: October 10, 2007
Creator: Aliberti, G. & Klann, R.
Partner: UNT Libraries Government Documents Department

The Effect of pH on Nickel Alloy SCC and Corrosion Performance

Description: Alloy X-750 condition HTH stress corrosion crack growth rate (SCCGR) tests have been conducted at 360 C (680 F) with 50 cc/kg hydrogen as a function of coolant pH. Results indicate no appreciable influence of pH on crack growth in the pH (at 360 C) range of {approx} 6.2 to 8.7, consistent with previous alloy 600 findings. These intermediate pH results suggest that pH is not a key variable which must be accounted for when modeling pressurized water reactor (PWR) primary water SCC. In this study, however, a nearly three fold reduction in X-750 crack growth rate was observed in reduced pH environments (pH 3.8 through HCl addition and pH 4-5.3 through H{sub 2}SO{sub 4} addition). Crack growth rates did not directly correlate with corrosion film thickness. In fact, 10x thicker corrosion films were observed in the reduced pH environments.
Date: October 10, 2002
Creator: Morton, D.S. & Hansen, M.
Partner: UNT Libraries Government Documents Department

Primary system boron dilution analysis

Description: The results are presented for an analysis conducted to determine the potential paths through which nonborated water or water with insufficient boron concentration might enter the LOFT primary coolant piping system or reactor vessel to cause dilution of the borated primary coolant water. No attempt was made in the course of this analysis to identify possible design modifications nor to suggest changes in administrative procedures or controls.
Date: October 10, 1978
Creator: Crump, R.J.; Naretto, C.J.; Borgen, R.A. & Rockhold, H.C.
Partner: UNT Libraries Government Documents Department

Heat transfer from internally-heated molten UO/sub 2/ pools. [PWR; BWR]

Description: Experimental measurements of heat transfer from internally heated pools of molten UO/sub 2/ have been obtained for two cell sizes: 10 cm x 10 cm and 20 cm x 20 cm. The experiments with the large cell have supported a previous conclusion from early small data that the measured downward heat fluxes are higher than would be expected on the basis of considerations of thermal convection. A convective model underpredicts the downward heat fluxes by a factor of 2.5 to 4.5 for all but one early experiment. Arbitrary assumptions of increased thermal conductivity do not account for the discrepancy. A single model based on internal thermal radiation heat transfer is able to account for the high values. The model uses the optically thick Rosseland approximation. Because of this, it is tentatively concluded that thermal radiation plays a dominant role in controlling the heat transfer from internally heated molted fuel.
Date: October 10, 1978
Creator: Stein, R.P.; Baker, L. Jr.; Gunther, W.H. & Cook, C.
Partner: UNT Libraries Government Documents Department