The comparison of available data on PWR assembly thermal behavior with analytical predictions

"PB-299 393." Topical report for Task #3 of the Nuclear Reactor Safety Research Program sponsored by New England Electric System, Northeast Utilities Service Co. under the MIT Energy Laboratory Electric Power Program.New England Electric System. Northeast Utilities Service Co

A review of recent analytical and experimental studies applicable to LMFBR fuel and blanket assembly design / by E. Khan and N. Todreas

"September, 1973."Includes bibliographical references (pages [40]-[44])U.S. Atomic Energy Commission contract AT(11-1)-224

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

Evaluation of the parfait blanket concept for fast breeder reactors

"January 1974."Also issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1974Includes bibliographical references (pages 261-264)An evaluation of the neutronic, thermal-hydraulic, mechanical and economic characteristics of fast breeder reactor configurations containing an internal blanket has been performed. This design, called the parfait blanket concept, employs a layer of axial blanket fuel pellets at the core midplane in the fuel pins of the inner enrichment zone; otherwise, the design is the same as that of the conventional LMFBR's to which the parfait configuration was compared. Two significant advantages were identified for the parfait blanket concept relative to the conventional design. First, the parfait configuration has a 25% smaller peak fast flux which reduces wrapper tube dilation by 37% and fuel element elongation by 29%; and second, axial and radial flux flattening contribute to a 7. 6% reduction in the peak fuel burnup. Both characteristics significantly diminish the problems of fuel and metal swelling. Other advantages identified for a typical parfait design include: a 25% reduction in the burnup reactivity swing, which reduces control rod requirements; a 7% greater overpower operating margin; an increased breeding ratio, which offsets the disadvantage of a higher critical mass; and more favorable sodium voiding characteristics which counteract the disadvantage of an 8% smaller power Doppler coefficient. All other characteristics investigated were found to differ insignificantly or slightly favor the parfait design.U.S. Atomic Energy Commission contract AT(11-1)-225

Development of a method for BWR subchannel analysis

Originally presented as the author's thesis, Ph.D. in the M.I.T. Dept. of Nuclear Engineering, 1980.This study deals with the development of a computer pro- gram for steady-state and transient BWR subchannel analysis. The conservation equations for the subchannels are obtained by area-averaging of the two-fluid model conservation equa- tions and reducing them to the drift-flux model formulation. The conservation equations are solved by a marching type technique which limits the code to analysis of operational transients only. The transfer of mass, momentum and energy between adjacent subchannels is split into diversion cross- flow and turbulent mixing components. The transfer of mass by turbulent mixing is assumed to occur in a volume-for- volume scheme reflecting experimental observations. The phenomenon of lateral vapor drift and mixing enhancement with flow regime are included in the mixing model of the program. The following experiments are used for the purpose of the assessment of the code under steady-state conditions: 1) GE Nine-Rod tests with radially uniform and nonuniform heating 2) Studsvik Nine-rod tests with strong radial power tilt 3) Ispra Sixteen-rod tests with radially uniform heating Comparison of calculated results with these data shows that the program is capable of predicting the correct trends in exit mass velocity and quality distributions

Analysis of mixing data relevant to wire wrapped fuel assembly thermal-hydraulic design

Statement of responsibility on title-page reads: E.U. Khan, N.E. Todreas, W.M. Rohsenow , and A.A. Sonin"September 1974."Includes bibliographical referencesIn this report analysis of recent experimental data is presented using the ENERGY code. A comparison of the accuracy of three types of experiments is also presented along with a discussion of uncertainties in utilizing this data for various code calibration purposes. The existence of internal swirl is discussed. The two empirical coefficients in ENERGY are determined from the data within a certain range of accuracy. This range is dictated to a large extent by the accuracy of the experiments and to a smaller extent by the ability of the code to utilize all sets of data in each experiment. The effect of geometry and bundle size on mixing and swirl flow is discussed. A realistic estimate of the degree of accuracy within which we can predict temperature distribution within the bundle and along the duct of a 217-pin wire wrapped fuel assembly of an LMFBR is presented. Gaps in data which need to be filled in to enhance our confidence in predicting coolant te! mperature distributions in a 217-pin LMFBR fuel bundle, are given. A brief description of two experiments that would fill these data gaps is presented. A novel experiment which would be very useful for both fuel and poison assembly mixing studies is described. Conclusions drawn from this study are believed to be quite general in nature.U.S. Atomic Energy Commission contract AT(11-1)-224