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1. |
Primary-Recoil Atom Spectra from ENDF/B Data |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 155-163
JenkinsJ. D.,
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摘要:
Information from the Evaluated Nuclear Data File (ENDF/B) has been used to construct energy-exchange kernels describing the energy transfer probabilities between neutrons and material lattice atoms. The kernels correctly incorporate all the available information on elastic scattering and inelastic scattering contained in ENDF/B. Their use removes much of the uncertainty in the calculation of primary-recoil spectra.Primary-recoil atom spectra for several elements in different reactor spectra have been generated. The influence of inelastic scattering on the recoil spectra is discussed.Results obtained with the ENDF/B data are compared with previously published information.
ISSN:0029-5639
DOI:10.13182/NSE70-A20703
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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2. |
Neutron Spectrum Measurements in a Homogeneous Erbium System and Erbium Cross Sections |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 164-176
GozaniT.,
TrimbleG. D.,
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摘要:
Careful measurements and calculations of the neutron spectrum in an aqueous solution of erbium nitrate are described. The validity of the infinite medium calculation is scrutinizingly studied. As a result it is shown that the discrepancy between measurement and calculation is caused by a too high (by about 10%) absorption cross section of erbium. This result is substantiated by two independent integral tests; the fundamental time eigenvalue of the system and the neutron conservation test.
ISSN:0029-5639
DOI:10.13182/NSE70-A20704
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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3. |
Neutron Fission Cross Sections for231Th,233Th,235U,237U,239U,241Pu, and243Pu from 0.5 to 2.25 MeV Using (t, pf) Reactions |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 177-187
CramerJ. D.,
BrittH. C.,
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摘要:
Experiments were performed using the (t, p) stripping reaction followed by fission to provide experimental measurements of fission probabilities. The neutron-induced fission cross section of the corresponding even-odd neutron targets can be deduced from these experimental (t, pf) data with the aid of an optical model calculation of the cross section for the formation of the compound nucleus by neutron absorption. Experimental measurements of the (n, f) cross section of the longer-lived targets such as235U and241Pu are used to test the validity of this technique by directly comparing experimental results and computed results from (t, pf) experiments. Ratios ofΓn/Γfare determined from these experimental data and are compared to previously published values.
ISSN:0029-5639
DOI:10.13182/NSE70-A20705
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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4. |
Total Neutron Cross Section of Technetium-99 from 0.01 to 1000 eV |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 188-192
WatanabeT.,
ReederS. D.,
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摘要:
The neutron transmission of99Tc for neutron energies less than 1 keV has been measured with the Material Testing Reactor (MTR) fast chopper with a resolution of 0.04 to 1.8µsec/meter. A total neutron cross section at 0.0253 eV of 24.7±1.7 b was obtained. To fit the cross-section data in the thermal energy range, it was necessary to assume a contribution by a bound level together with contributions from measured resonances at positive energies. Resonance parameters are presented for levels observed in the energy region from 0.01 to 300 eV. Two additional resonances, not listed in the literature, have been measured and analyzed. Parameters of individual resonances below 300 eV and average parameters at higher energies, give a resonance absorption integral of 340±20 b, and a value of (0.43±0.14)×10−4for thes-wave neutron strength function.
ISSN:0029-5639
DOI:10.13182/NSE70-A20706
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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5. |
A Unified Model of Deformed Odd-Odd Nuclei for Nuclear Data Generation and Analysis |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 193-208
FuC. Y.,
YostK. J.,
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摘要:
A unified model of deformed odd-odd nuclei has been formulated as an aid in nuclear data generation and evaluation. The model employs products of single-particle Nilsson wave functions as basis functions. The coupling of angular momenta of the odd nucleons is assumed to obey the Gallagher-Moszkowski coupling rules. The matrix elements of the proton-neutron residual interaction potential are evaluated with the use of oscillator brackets. The validity of the model has been established by computing and comparing with experimental data nuclear-energy levels and/or gamma-ray transition probabilities for23Na,28Al,166Ho,182Ta, and238Np. The calculated results compare quite well with experiment. Special attention has been given to the establishment of an efficient computational method.
ISSN:0029-5639
DOI:10.13182/NSE70-A20707
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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6. |
Reaction Rates in Representative Plutonium-Recycle Fuel Lattices |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 209-225
CarverJ. G.,
MorganW. R.,
PorterC. R.,
RobkinM. A.,
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摘要:
Measurements have been made of relative nuclear-reaction rates within sub-critical water-moderated plutonia-urania fuel lattices, under conditions considered typical for plutonium recycle in central-station power reactors. Measurement conditions included water:fuel ratios of 3:1 and 2:1; temperatures of 70, 235, 330, 430, and 540°F; and three positions within the unit cell. Nuclear reaction rates measured included relative fission rates in235U,239Pu, and241Pu, as well as relative capture rates in176Lu (principal resonance at 0.143 eV),168Yb (0.597 eV),191Ir (0.654 eV),193Ir (1.303 eV),197Au (4.906 eV),139La (73.5 eV), and63Cu (1/vdetector). To facilitate comparison with predicted values, the experimental resonance absorption-rate ratios were normalized to ratios measured within a pure water spectrum. Experimental reaction-rate ratios were compared with values predicted using the THERMØS code in conjunction with a modified version of EPITHERMØS; and agreement varying from fair to good was observed. The internal consistency of the measurements suggests their future utility for evaluating methods of calculating neutron spectra and relative reaction rates within lattices of the type considered.
ISSN:0029-5639
DOI:10.13182/NSE70-A20708
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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7. |
Linearly Anisotropic Extensions of Asymptotic Neutron Diffusion Theory |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 226-236
DoyasR. J.,
KoponenB. L.,
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摘要:
In this paper, anisotropic transport-theory corrections to asymptotic diffusion theory are suggested and evaluated. These modifications are found to provide a natural and useful extension of diffusion-theory applications in one-dimensional, one-region problems.An effort is also made to determine whether the treatment of problems of more than one dimension in one-dimensional codes can be improved through the use of this extended formulation of diffusion theory.
ISSN:0029-5639
DOI:10.13182/NSE70-A20709
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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8. |
Truncation Error Analysis of Finite Difference Approximations to the Transport Equation |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 237-248
ReedWm. H.,
LathropK. D.,
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摘要:
The truncation error of difference approximations to the transport equation is examined, and difference equations which are uniformly second-order accurate are derived. Resulting angular quadrature sets are shown to integrate exactly polynomials of second order in the direction cosines. The analysis is completed for the most general cases of three dimensional spherical and cylindrical geometries. Numerical results comparing this second-order scheme to the standard diamond difference equations in one-dimensional spherical geometry are presented.
ISSN:0029-5639
DOI:10.13182/NSE70-A20710
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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9. |
Spectral Synthesis Applied to Fast-Reactor Dynamics |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 249-258
StaceyWeston M.,
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摘要:
Spectral synthesis methods, in which the neutron flux spectrum is synthesized from known trial spectra, is applied to fast-reactor dynamics. Numerical results demonstrate that important spectral shift effects which occur during transients are adequately represented by synthesis models with two trial spectra. Conventional few-group models which require equivalent computational effort are found to provide a significantly inferior representation of these effects.
ISSN:0029-5639
DOI:10.13182/NSE70-A20711
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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10. |
Monte Carlo Method for Geometrical Perturbation and its Application to the Pulsed Fast Reactor |
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Nuclear Science and Engineering,
Volume 41,
Issue 2,
1970,
Page 259-270
TakahashiHiroshi,
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PDF (897KB)
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摘要:
The Monte Carlo method is applied to calculate the reactivity change due to a moving reflector block in the pulsed fast reactor system. This reactivity change is an important quantity in the determination of the power pulse width. In the important region of small displacements about the maximum reactivity point, the reactivity change is so small that the ordinary Monte Carlo methods, using the importance sampling, Russian roulette, or splitting techniques, require prohibitively long calculation times. To avoid this difficulty, a new Monte Carlo method, which directly calculates the geometry coefficient of reactivity due to a geometry perturbation, is developed by adopting the method used in the calculation for the Doppler coefficient by Olhoeft. The formulation of the new method is discussed. The GEMCM code for this geometry perturbation, which is made by modifying the 05R code, is described. Finally, an analysis of the critical experiment for the pulsed fast reactor is carried out using this method and the applicability of the method is discussed.
ISSN:0029-5639
DOI:10.13182/NSE70-A20712
出版商:Taylor&Francis
年代:1970
数据来源: Taylor
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