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Description: jectives, the accomplishments, and a summary of the work outstanding. The obtectives of the experimental and analytical studies were to investigate and reach an understanding of the kinetic behavior of aqueous homogeneous reactors. Information produced by the program, experiments on the spherical core, capsule experiments, and the remaining work schedule are discussed. (W.D.M.)
Date: January 28, 1959
Creator: Flora, J.W.
Partner: UNT Libraries Government Documents Department


Description: plants ln England and France. With the increasing de output of given designs and probably allow operation at higher polymer contents than orignally foreseen, thereby reducing the make-up requirements. The physical characteristics of the OMRE such as critical loading, temperature coefficient, and general stability appeared to be close to the predicted values. Radiation levels in the primary circuit area during full power operation appear to be so low that maintenance is possible during operation. The reactor has been run for a full month at 30% polymer concentration and is, at the time of this writing, brought to a still higher steady state percentage of breakdown products ln the coolant stream. No evidence whatsoever of fouling or precipitation has been observed. The reactor behaves in a routine manner in all respects and invites immediate application of the OMR principle to reactors for large scale ceniral stations. Final design on one 11.4 Mwe unit for the city of Piqua, Ohio, has now stanted. A short description is given of OMR power reactors. The use of magnetic jack mechanisms for control and safety rods provides a reactors top shield without penetrations, as well as an unpenetrated lower core vessel, still avoiding any interference from the control rods during fuel changing. The new finned-plate fuel element is mentioned as well as the use of a liquid pressurizing pump instead of nitrogen gas pressurization. It is conservatively predicted that the cost of organic liquid make-up for these designs will not contribute more than one half to one mill per kwh to the total power cost. In case operation at higher polymer concentrations appears practicable, this figure may even be lower. More detailed pricing informntion available now, has shown that the original cost estimate of around 0 per kw installed for a 150 Mwe plant can ...
Date: September 20, 1958
Creator: Garbe, R.W. & Walchli, H.E.
Partner: UNT Libraries Government Documents Department


Description: BS>This work is based on a rather specific and simplified model of the reactor core in which internal circulation exists. Because of the model simplicity, the influence varying some of the parameters appears more directly. In particular, calculation of the circulation factor, which is important in determining the need for vapor separators is included. On this basis it was felt desirable to make the results of this study available even though they have been largely superseded by the later work. (auth)
Date: October 1, 1952
Creator: Zmola, P. C. & Lawson, E. N.
Partner: UNT Libraries Government Documents Department

Iodine transport analysis in the ESBWR.

Description: A simplified ESBWR MELCOR model was developed to track the transport of iodine released from damaged reactor fuel in a hypothesized core damage accident. To account for the effects of iodine pool chemistry, radiolysis of air and cable insulation, and surface coatings (i.e., paint) the iodine pool model in MELCOR was activated. Modifications were made to MELCOR to add sodium pentaborate as a buffer in the iodine pool chemistry model. An issue of specific interest was whether iodine vapor removed from the drywell vapor space by the PCCS heat exchangers would be sequestered in water pools or if it would be rereleased as vapor back into the drywell. As iodine vapor is not included in the deposition models for diffusiophoresis or thermophoresis in current version of MELCOR, a parametric study was conducted to evaluate the impact of a range of iodine removal coefficients in the PCCS heat exchangers. The study found that higher removal coefficients resulted in a lower mass of iodine vapor in the drywell vapor space.
Date: March 1, 2009
Creator: Kalinich, Donald A.; Gauntt, Randall O.; Young, Michael Francis & Longmire, Pamela
Partner: UNT Libraries Government Documents Department

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.
Date: December 31, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. & Xiang, J.Y.
Partner: UNT Libraries Government Documents Department


Description: The analysis of the radiation sources within the primary concrete shielding of the Yankee Atomic Electric Plant is summarized The dose due to these sources at a point on a transverse centerline and outside the plant container was calculated. The dose at two points on the axial centerline was found as well as the extent to which the water is actiwated during its passage through the pressure vessel. (auth)
Date: October 1, 1957
Creator: Graves, H.W. Jr.; Eich, W.J. & Williams, H.T. Jr.
Partner: UNT Libraries Government Documents Department


Description: BS>The calculation of control rod worths in light water moderated reactors is presented. Considerable experimental information from the Yankee critical experiments is discussed and applied to improve the theory. The calculation system thus deduced from theory and experiment is then applied to the final design of the first Yankee core to provide a prediction of the shutdown available in its control rods. These calculations show that the 24 control rods presently provided are sufficient to shut the hot, clean, zero power first core down by 3%. (auth)
Date: September 1, 1959
Creator: Arnold, W.H. Jr.
Partner: UNT Libraries Government Documents Department


Description: The nuclear characteristics of the CETR are described. Core operating lifetime, control-rod worth, and powerdensity distribution are discussed in relation to maximizing the core operating life. Other objectives of nuclear design are to minimize the power-density variation and to assure control of the reactor. (J.R.D.)
Date: March 1, 1960
Creator: Barringer, H.S.; Flickinger, R.B. & Spetz, S.W.
Partner: UNT Libraries Government Documents Department


Description: An attempt is made to set forth in systematic form the two-energy group physics calculations for the Argonaut one-slab configuration. All assumptions are stated and full calculational details given so that the procedure used may be followed. Complete point by point flux values from the PDQ programming of the problem on the IBM-704 are given. A comparison of theoretical and experimental results is included. (W.D.M.)
Date: April 1, 1960
Creator: Moon, D.P.
Partner: UNT Libraries Government Documents Department


Description: A description of the nuclear and thermal analysis involved in the design of the WCAP-4 in-pile test is presented. Calculations reveal that no burn-out conditions are expected at the operating fluxes. Calculations of central core temperature and heat output are included. Various hot spot factor calculations for both non-local and local boiling conditions are also given. (J.R.D)
Date: January 1, 1960
Creator: Bourina, A.; Eich, W. & Lombardo, J.J.
Partner: UNT Libraries Government Documents Department


Description: The research and development program pertinent to the conceptual design and ultimate construction at ANL of an advanced research reactor with a peak thermal flux of 5 x 10/sup 15/ n/cm/sup 2//sec is documented. The basic reactor complex, the problems involved, the various approaches pursued, the present status and estimated cost of the project, along with recommendations for future research and development essential to the successful culmination of the project are described. The reactor is moderated with D,Oand has a core life of 120 hours at 250 Mw, (W.D.M.)
Date: September 1, 1959
Creator: Link, L.E.; Armstrong, R.H.; Cameron, T.C.; Dickson, R.F.; Heineman, J.B.; Kelber, C.N. et al.
Partner: UNT Libraries Government Documents Department


Description: An attempt is made to provide a single reference source for material pertinent to the maintenance and operation of the AE-6 Reactor. Descriptions of various components are included, as well as the common operational and maintenance procedures and check lists. Information is given on the subcritical assemblies that may be studied. (W.D.M.)
Date: June 23, 1960
Creator: Moore, E.A. Jr.
Partner: UNT Libraries Government Documents Department