摘要:

AbstractAn indirect experimental technique for determining the expansion coefficient was developed to provide uncertainty estimates for expansion coefficient calculations. This technique uses an R, Z reactivity worth map synthesized from small-sample reactivity traverse measurements for major materials over the reactor core and blanket regions. The experimentally based expansion coefficients, representing the reactivity change due to uniform axial and radial expansion, are deduced by appropriately integrating measured worth profiles. This technique was evaluated in Phase A of the Zero Power Plutonium Reactor Assembly 5. Direct calculations of the expansion coefficients were performed, and results were compared with the experimentally determined values. The validity of the technique used to derive expansion coefficients from worth measurements was evaluated.It is concluded that the total expansion coefficients are reasonably well calculated; however, the calculated radial expansion coefficient was overestimated. Sources of possible systematic errors in the experimentally based values were studied. Based on the present experiment, an uncertainty of±20% (90% level of confidence) on expansion calculations using ENDF/B-III data is estimated for a clean core configuration.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27174

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractEleven233U solution critical assemblies spanning an H/233U ratio range of 40 to 2000 and an unreflected metal233U assembly have been calculated with ENDF/B-IV and Hansen-Roach cross sections. Results from these calculations are compared with the experimental results and with each other. We observed an increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/233U ratio, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range, whereas the ENDF/B calculations yielded high values for keffon assemblies of low moderation.We conclude that serious problems exist in the ENDF/B-IV representation of the233U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, we recommend that an experimental program be undertaken to obtain233U criticality data at low H/233U ratios for verification of generalized criticality safety guidelines.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27175

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractNeutron capture and fission cross sections of241Pu have been measured from 0.01 eV to 30 keV, and their ratio has been measured up to 250 keV. The cross sections were normalized at thermal-neutron energies (0.02 to 0.03 eV) to the ENDF/B-IV evaluation. The source of pulsed neutrons was the Oak Ridge Electron Linear Accelerator. The gamma-ray detector used to detect capture and fission events was the“total energy detector,”which is a low-efficiency detector whose average efficiency is forced to be proportional to the energy of the interacting gamma rays by weighting these events according to their pulse height in the scintillator. Fast-neutron scintillation detectors with pulse-shape discrimination were used to detect fission events. The shape of the neutron flux was measured relative to the10B(n,α) cross section. The measurements are unique for241Pu in that absorption and fission were determined directly and simultaneously over a wide neutron energy range rather than indirectly by inferring capture from separate fission and total cross-section measurements. The results indicate that the neutron resonance region of the ENDF/B-IV evaluation underestimates capture by a factor of∼2. Above the resonance region (∼100 eV), there are no previous measurements of the differential capture cross section. These cross sections are important in plutonium-fueled reactors.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27176

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractPredicted fission cross-section values, based on systematic trends observed recently by Behrens, are presented in the neutron energy interval from 3 to 5 MeV for 43 isotopes, ranging from thorium through californium. Comparisons with measured fission cross-section values are included for 33 isotopes.

ISSN:0029-5639    

DOI:10.13182/NSE65-464

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractThe currently used formalism for neutron cross-section representation in the unresolved resonance energy range is based on the statistical parameters of the population of Breit-Wigner resonances. The present work introduces practical formalisms, based on parametric representation of the shielding factor curves, by which the values of effective cross sections can be obtained simply and quickly in the unresolved range, and suggests their use for neutron data representation. These formalisms were found to be compatible with such existing codes as MC2, ETOX, HAMMER, ENDRUN, and MIGROS, and with such existing nuclear data files as ENDF/B and KEDAK.Each formalism is based on one interpolation scheme in temperature and one inσ0. The accuracy of four schemes in temperature and three schemes inσ0was checked. Of these, three temperature schemes and oneσ0scheme were found to have better than 1% accuracy in the entire unresolved region, thus yielding a formalism with better than 2% accuracy for representation.Observed spatially dependent self-shielding factors are transformed into pseudo-background cross-section-dependent (Bondarenko-type) self-shielding factors. Numerical values of the transformation for235U and239Pu self-shielding factors are given. It is shown that the formalisms can be used for the preprocessing of current nuclear data files in the unresolved range.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27178

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractBenchmark experiments of neutron propagation in iron and iron-sodium mixtures were used to generate an“adjusted”ENDF/B data file for iron, Mat = 1192. In particular, the secondary neutron energy distribution in the continuous level energy range was adjusted using such high-energy responses as the32S(n,p)32P reaction, which are significantly sensitive to changes in that probability distribution. The experimental analysis used carefully checked two-dimensional transport methods to avoid bias in the adjustment procedure due to inadequate calculational methods.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27179

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractMaterial bucklings have been determined as functions of235U enrichment in UO2(0.7, 1.2, and 1.5 wt%235U), PuO2enrichment in PuO2-UO2(0.54 and 0.87 wt% PuO2), fissile content of plutonium (91 and 75% Pu-fissile), lattice pitch (Vmod/Vfuel: 7.4 and 9.9), and coolant void fraction. The reference loading of 1.2 wt%235U-enriched UO2clusters was progressively replaced by the test clusters.Buckling differences resulting from the substitutions were analyzed by the new second-order (iterative) perturbation method, on the assumption that neutron diffusion is isotropic and that no difference in diffusion coefficients exists between the two lattices. This analysis takes into account the effect of distortion in radial neutron flux distribution in the substituted core without any iterative correction procedure that is usually adopted in the first-order perturbation method. Also, it is not necessary in the case of the present analysis to introduce any usual intermediate region for taking into account the effect of spectrum mismatch between the two lattices. The material buckling differences between the test and reference lattices, which are in the range of−10.2 to 9.1 m−2, were determined within 3% of uncertainty.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27180

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractBased on the one-delay-group point reactor model, the influence of the transport time delay on the nature of the linear stability of reactor dynamics is studied with the aid of the method of D partitions. From our analysis, the stability domain can be easily determined and plotted in the parametric space. The domain of the linear stability is significantly altered by the delayed temperature feedback. Comparing the stability domain of the one-group model with the effective lifetime model and Welton's criterion, we obtain the following conclusions:1.The straight line obtained from Welton's criterion is a tangent line of the dynamic stability boundary of the effective lifetime model.2.The effective lifetime model is a safe estimation of the linear stability only when the delayed neutron precursor decay constant is greater than the reciprocal time constant for heat removal.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27181

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractThe CNmethod of solving the transport equation is applied to one-velocity classical problems in cylindrical geometry. The cylindrical angular Green's function is calculated first, the inner and outer albedo problems are solved, and numerical results are obtained and compared with reference results. The extrapolation length for the cylindrical Milne problem is calculated, and accurate results are presented. Finally, the determination of the critical radius for multiplying media is treated.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27182

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor

摘要:

AbstractIn the framework of the solution of the transport equation by means of diffusion techniques, an iterative procedure is presented that permits us to calculate the point-dependent diffusion coefficient, Dk(k = x,y,z), using standard diffusion codes. Numerical comparisons show that this procedure attains a flux distribution much closer to the transport distribution after one iteration than the classical diffusion flux. The time of the calculation is about twice that required by classical diffusion.

ISSN:0029-5639    

DOI:10.13182/NSE78-A27183

出版商:Taylor&Francis    

年代:1978

数据来源: Taylor