1. |
Eta of U233from 1 to 800 ev* |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 105-119
YeaterM. L.,
HockenburyR. W.,
FullwoodR. R.,
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摘要:
Eta of U233has been measured in the energy range from 1 to 800 ev by counting fast fission neutrons in ZnS-loaded plastic scintillators; incident neutron energies were measured in a time-of-flight system with a betatron as pulsed source. The data are normalized at one electron volt to data reported at Geneva by the MTR group; the latter results were normalized to an absolute thermal value. Corrections have been made for multiple scattering by a rigorous single velocity analysis and by a Monte Carlo calculation which also accounts for energy degradations.
ISSN:0029-5639
DOI:10.13182/NSE61-A15594
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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2. |
Cold Neutron Beams from Small Low-Temperature Moderators in Reactors |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 120-126
WebbF. J.,
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摘要:
An investigation has been made of the efficiency of several hydrogenous low-temperature moderators in the production of intense cold neutron beams from reactors. The dimensions of such a moderator are usually severely limited, but even small volumes of liquid hydrogen, liquid hydrogen deutride, and solid methane were effective. A qualitative theoretical explanation is given of the degree of moderation observed.
ISSN:0029-5639
DOI:10.13182/NSE61-A15595
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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3. |
Error Propagation in Hot-Spot, Hot-Channel Analysis* |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 127-131
TingeyFred H.,
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摘要:
The use of hot-spot, hot-channel factors to relate system uncertainties and potential system temperatures in reactor design is fairly standard. In this paper a somewhat different approach is discussed. Translating factors are identified for each base variable in the calculation and through statistical regression analysis and error synthesis, procedures are outlined for approximating to the hot-spot temperature and estimating the associated uncertainty in the calculation. The procedures are applied to a typical ETR calculation.
ISSN:0029-5639
DOI:10.13182/NSE61-A15596
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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4. |
Thermal Neutron Absorption Cross Sections by the Pulsed Source Method* |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 132-136
MeadowsJ. W.,
WhalenJ. F.,
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摘要:
The thermal neutron absorption cross sections of twenty-one naturally occurring elements and B10have been determined by comparing the time dependence of the neutron flux in water with the time dependence of the neutron flux in a water solution of the sample with the same geometric buckling. After making some small corrections arising largely from the change in the number of hydrogen atoms per cm3in the solution, the decay constant of the ratio curve gives the macroscopic absorption cross section averaged over the neutron flux spectrum. For a 1/υcross section the 2200 m/sec cross section can be directly computed. For non-1/υcross sections the effective 2200 m/sec cross section is obtained.
ISSN:0029-5639
DOI:10.13182/NSE61-A15597
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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5. |
Radiation Damage in Steel: Considerations Involving the Effect of Neutron Spectra |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 137-147
RossinA. D.,
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摘要:
The mechanism of interaction between fast neutrons and atoms of a metal lattice is described. A cross section for the production of vacancies in iron by neutrons, as a function of neutron energy, is derived and shown to be roughly proportional to the product of the neutron energy and the isotropic elastic scattering cross section. The vacancy production cross section is applied to several reactor spectra and the results show that an appreciable fraction of the radiation damage in crystalline solids, particularly metals, can be caused by neutrons having energies below 1 Mev. Also the assumption that the neutrons responsible for radiation damage have a fission spectrum distribution appears to be inapplicable in reactor situations. In fact, no quantitative measure of total neutron exposure can be made without knowledge of the spectral shape. Steel is chosen as an example because of the interest in its properties as a function of irradiation, hence the model is developed based on interaction of neutrons with iron atoms. Some important limitations of the method are cited.
ISSN:0029-5639
DOI:10.13182/NSE61-A15598
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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6. |
Gamma Branching in Kr85 |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 148-150
LyonW. S.,
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摘要:
Kr85decays directly to the ground state by 0.67-Mev maximum beta emission in greater than 99% of the decay. The fraction of the decay which proceeds through a lower energy beta emission (∼0.16 Mev) followed by a gamma transition of 0.514 Mev has been determined by measurement of the number of 0.513-Mev gamma rays per beta disintegration. Thisγ/βratio measured byγscintillation spectrometry and absolute beta counting techniques was found to be 0.0038±0.0003.
ISSN:0029-5639
DOI:10.13182/NSE61-A15599
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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7. |
Comparison Theorems for the Estimation of Criticality in Bare, One-Velocity Reactors |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 151-156
LawrenceDresner,
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摘要:
Some theorems are given for the estimation of the criticality of bare, one-velocity reactors of irregular shape by comparison with similar reactors of regular shape for which the critical parameters are known.
ISSN:0029-5639
DOI:10.13182/NSE61-A15600
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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8. |
Neutron Thermalization and Diffusion in Pulsed Media* |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 157-167
PurohitS. N.,
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摘要:
A general formalism for determining the lower time eigenvalues associated with a decaying pulse of neutrons in a finite multiplying as well as nonmultiplying medium has been developed. This formalism is based upon the expansion of each energy eigenfunction by a complete sum of the associated Laguerre polynomials of first order. The eigenvalues are expressed in terms of the energy transfer moments of the scattering kernel of the medium, weighted by the Maxwellian distribution. The importance of the first eigenvalue in the establishment of the final asymptotic energy distribution is discussed. In the case of a nonabsorbing infinite medium, the reciprocal of the first eigenvalue is shown to be equal to the thermalization time constant, with which the Maxwellian velocity distribution of neutrons is attained. The thermalization time constant was estimated for various moderators. For the heavy-gas case, the thermalization time constant was was found to be equal to (1.274°ζ∑s0υ0)−1. It is also established in this study that only two polynomials are required to obtain the relation between the thermalization time constant and the diffusion cooling coefficient derived previously from the Rayleigh-Ritz variational principle. The formalism presented in this paper is general and avoids the concept of neutron temperature in defining the thermalization time constant. The decay of a neutron pulse in a nonmultiplying medium is discussed in detail. For the case of multiplying medium, an analysis of an experiment is presented to indicate the importance of the time-dependent nonleakage probability in the expression of the zeroth eigenvalue.
ISSN:0029-5639
DOI:10.13182/NSE61-A15601
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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9. |
Measurements of Neutron Spectra in Water, Polyethylene, and Zirconium Hydride* |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 168-184
BeysterJ. R.,
WoodJ. L.,
LopezW. M.,
WaltonR. B.,
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摘要:
An experimental arrangement designed for accurate measurements of low-energy neutron spectra has been assembled and tested.A pulsed high-current electron linear accelerator is used to produce short bursts of fast neutrons which are introduced into a moderating and absorbing assembly. The steady-state energy spectrum of neutrons in the assembly is determined by pulsed-beam time-of-flight techniques. Hydrogen-moderated systems poisoned with a number of common neutron absorbers (boron, cadmium, samarium) have been studied, and the resulting spectra compared with theoretical predictions using both free and bound hydrogen scattering kernels. In general, a marked difference exists between measured spectra and spectra calculated using a free hydrogen kernel. In the case of water where a detailed scattering kernel is available for room temperature, theory and experiment are in reasonable agreement.
ISSN:0029-5639
DOI:10.13182/NSE61-A15602
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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10. |
On the Solution of Transport Problems by Conditional Monte Carlo |
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Nuclear Science and Engineering,
Volume 9,
Issue 2,
1961,
Page 185-197
DrawbaughD. W.,
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摘要:
The application of a conditional Monte Carlo method for estimating the solution of transport problems is described. Statistical estimators for solving transport problems by the conditional Monte Carlo method are given for both homogeneous and heterogeneous problems. The analog and conditional Monte Carlo solutions of a simple one-speed transport equation are compared with the exact deterministic solution. The application to deep penetration shielding problems is discussed.
ISSN:0029-5639
DOI:10.13182/NSE61-A15603
出版商:Taylor&Francis
年代:1961
数据来源: Taylor
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