|
1. |
Heat Transfer to Mercury Flowing In-Line Through an Unbaffled Rod Bundle: Effect of Rod Displacement on Local Surface Temperature and Local Heat Flux |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 321-335
DwyerO. E.,
HlavacP. J.,
HelfantM. A.,
Preview
|
PDF (2184KB)
|
|
摘要:
AbstractAn experimental study of heat transfer to mercury flowing longitudinally through an unbaffled rod bundle was carried out. The purpose was to determine the effect of lateral displacement of a rod on local heat transfer behavior. In a previous paper, the effects of extent and direction of displacement on the rod-average heat transfer coefficient were presented for the displaced rod, on that (or those) toward which it was displaced, and on that (or those) from which it was displaced. Here, the effects of extent and direction of displacement on the peripherally local heating surface temperature, local heat flux, local heat transfer coefficient, and local surface temperature fluctuations are presented for the displaced rod.The test bundle had aP/Dratio of 1.75, and the rods were special electrical heaters. It was found that rod displacement can cause a large circumferential variation in its local heat transfer characteristics. Aside from theP/Dratio, the independent parameters affecting these characteristics are circumferential angle (θ), relative cladding thickness [(r2−r1)/r2], relative cladding conductivity (kw/kf), and flow rate (Pe).It was found that displacement of a rod can produce circumferential variations in its surface temperature comparable to the average temperature drop from the heating surface to the coolant stream. For a given displacement, this variation increases as average heat fluxincreases and as (r2−r1)/r2,kw/kf, and Pe decrease; changes inhave the greatest effect, and those in (r2−r1)/r2andkw/kf, the least. For a given displacement and flow rate, the greater the surface temperature variation, the less will be the circumferential variation in the local heat flux. Thus, as either cladding thickness or conductivity increase, the variation in the local heat transfer coefficient (and therefore the average) remains about the same. It was found that, as a rod is displaced from its symmetrical position, the local heat transfer coefficients surprisingly decrease at all circumferential points, which partly explains why the rod-average heat transfer coefficient is highly adversely affected by lateral rod displacement. This is only true for liquid-metal coolants.
ISSN:0029-5639
DOI:10.13182/NSE70-A19090
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
2. |
The Two-Phase Critical Discharge of Initially Saturated or Subcooled Liquid |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 336-342
HenryRobert E.,
Preview
|
PDF (833KB)
|
|
摘要:
AbstractA nonequilibrium model is developed to describe the two-phase critical discharge of initially saturated and subcooled liquid through sharp edged inlet, constant area ducts withL/D≥12. A similar model is proposed for smooth inlet geometries withL/D≥0. The proposed solutions, which are based on the upstream stagnation conditions, exhibit excellent agreement with the published data which include water, Freon-11, and Freon-12 results.
ISSN:0029-5639
DOI:10.13182/NSE70-A19091
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
3. |
Statistical Evaluation of the Maximum Temperatures in Reactor Cores |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 343-350
AmendolaA.,
Preview
|
PDF (1065KB)
|
|
摘要:
AbstractA new statistical method for evaluation of hot channel and hot spot factors is presented. A new definition of“hot spot”is proposed with which the probability of exceeding critical temperatures can be correlated to the size of the zone in which they occur. In contrast to previous methods, the hot channel factors are demonstrated to be independent of the assumed spot size, provided that the uncertainties are correctly specified. Therefore, a new criterion is proposed for specification of the uncertainties which are random variables along the fuel pin axis, and the concept of a“specific standard deviation”is introduced. The different effects of the uncertainties, whether they act on single elements of the core, on groups of elements or on the whole core, are taken into account by an appropriate procedure. The statistical analysis takes into account the whole core with its particular axial and radial nominal temperature profiles. The principal results obtained by the SHØSPA code for the sodium-cooled fast reactor Na-2 are discussed.
ISSN:0029-5639
DOI:10.13182/NSE70-A19092
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
4. |
The Total Neutron Cross Section of Lithium-7 and Carbon from 100 to 1500 keV |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 351-356
MeadowsJ. W.,
WhalenJ. F.,
Preview
|
PDF (814KB)
|
|
摘要:
AbstractA precise determination of the neutron total cross section of7Li and carbon has been made in the energy region 100 to 1500 keV. The parameters of the prominent7Li resonance in the laboratory system areEλ= 261 keV,Γλ= 36.5 keV and= 594 keV. Corresponding parameters for the principal bound state resonance in carbon areEλ=−2020 keV and= 540 keV. The carbon data are fitted byσT= 4.830−3.55E+ 1.587E2−0.305E3, whereσTis in barns andEis in MeV. Above 500 keV the7Li data are fitted byσT= 6.929−27.018E+ 42.721E2−27.210E3+ 6.139E4.
ISSN:0029-5639
DOI:10.13182/NSE70-A19093
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
5. |
Measurements of Thermal Neutron Spectra in Heterogeneous Lattices by Foil Activation |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 357-366
DixonG. W.,
SherR.,
Preview
|
PDF (691KB)
|
|
摘要:
AbstractThermal neutron spectra have been measured with good spatial resolution within a unit cell in several H2O-moderated natural uranium lattices and in one graphite-moderated lattice. The H2O-moderated lattices had water-to-uranium volume ratios of 1:1, 2:1, and 3:1, with fuel rod diameters of 2.54 cm. Dysprosium-164,151Eu,176Lu, and115In were used as detectors, and both activation ratios and unfolded spectra are compared with THERMOS code calculations. The agreement between the results and the calculations is satisfactory; however, the agreement in the water regions is generally much better than in the fuel regions of the H2O-moderated lattices. In the graphite lattice, the agreement of results with THERMOS calculations using a free gas kernel is poor, while calculations with a crystalline kernel show better agreement.
ISSN:0029-5639
DOI:10.13182/NSE70-A19094
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
6. |
Comparison of Calculated and Measured Energy-Dependent Activation Rates for NERVA-Type Reactor |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 367-380
SalahS.,
RankinW. D.,
OblockV. S.,
Preview
|
PDF (12338KB)
|
|
摘要:
AbstractAbsolute reaction rates of thermal, resonance, and threshold detectors were calculated and measured within a graphite moderated, graphite reflected critical assembly using NERVA-type fuel elements to provide verification of analytical techniques and basic neutron cross-section data. Detectors used were: Dy, In, Au, W, Mn, Cu,235U, and238U foils and S pellets. Comparison of the calculated energy-dependent reaction rates with measured values showed them to be generally within experimental uncertainties. Near the outer edge of the reactor, however, the difference between the calculated and experimental values is greater than the experimental uncertainties. The comparison of these calculations and measurements show that the spatially dependent neutron spectra are adequately predicted with the multigroup, multiregion transport calculations utilized in this analysis.
ISSN:0029-5639
DOI:10.13182/NSE70-A19095
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
7. |
Continuous Slowing Down Theory Applied to Fast-Reactor Assemblies |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 381-393
StaceyWeston M.,
Preview
|
PDF (8293KB)
|
|
摘要:
AbstractContinuous slowing down theory is applied to the treatment of the elastic moderation of neutrons in a fast-reactor assembly, where strong scattering resonances are prevalent. Three prescriptions for the moderating parameters, one of which is original to this paper, are studied. Continuous slowing down theory is demonstrated to yield quite accurate results for the neutron spectrum, when the improved prescriptions presented in this paper are used. A method for treating narrow resonance absorption separate from the slowing down calculation, then using the results of the former to attenuate the flux obtained from the latter, is presented and demonstrated to be promising.
ISSN:0029-5639
DOI:10.13182/NSE70-A19096
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
8. |
An Exact Solution of The Neutron Slowing Down Equation |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 394-398
StefanovićD.,
Preview
|
PDF (335KB)
|
|
摘要:
AbstractThe slowing down equation for an infinite homogeneous monoatomic medium is solved exactly. The cross sections depend on neutron energy. The solution is given in analytical form within each of the lethargy intervals. This analytical form is the sum of probabilities which are given by the Green functions. The calculated collision density is compared with the one obtained by Bednarz and also with an approximate Wigner formula for the case of a resonance not wider than one collision interval. For the special case of hydrogen, the present solution reduces to Bethe's solution.
ISSN:0029-5639
DOI:10.13182/NSE70-A19097
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
9. |
Higher Order Variational Principles and Iterative Processes |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 399-403
DevooghtJ.,
Preview
|
PDF (315KB)
|
|
摘要:
AbstractN-step iterative processes give rise toN-order variational principles. Thep-hyperpower iterative process of Altman and Petryshyn is shown to be equivalent to thep-order variational principle of Kostin and Brooks. A higher order Chebychev variational principle is derived, which is independent of the normalization of the trial operator.
ISSN:0029-5639
DOI:10.13182/NSE70-A19098
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
10. |
Time-Dependent Neutron Transport Theory for Fast Pulsed Assemblies |
|
Nuclear Science and Engineering,
Volume 41,
Issue 3,
1970,
Page 404-416
KnickleH. N.,
DaitchP. B.,
Preview
|
PDF (1007KB)
|
|
摘要:
AbstractThe time-dependent transport equation in plane geometry has been solved numerically using the double spherical harmonics angular approximation and first-order finite differences. The monoenergetic case has been shown to meet both necessary and sufficient conditions for stability for reasonable values of the time step. The convergence and wave front propagation characteristics of the difference scheme have also been checked in special cases and found to be satisfactory. A computer program has been written to solve the difference equations of the multienergy, multiregion problem. Monoenergetic and multigroup calculations have been made which compare qualitatively with experimental results.
ISSN:0029-5639
DOI:10.13182/NSE70-A19099
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
|
|