GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra. [In FORTRAN for CDC 7600]

Young, P.G.; Arthur, E.D.

doi:10.2172/5224192

Cited by 53 publications

(39 citation statements)

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“…1 15) on the total cross section of 56 Fe. The resonance formulas for 57,58 Fe were changed from the Breit-Wigner type to the Reich-Moore type, although the parameter values remain unchanged. Figure 2 shows the total cross section of vanadium.…”

Section: Evaluation 1 Light and Medium-heavy Nuclide Datamentioning

confidence: 99%

“…In principle, a weighted sum of the evaluated total cross section of an isotope should re- flect the measured total cross section of the element. For instance, in the MeV region, the total cross sections of 57,58 Fe were obtained from the optical model calculations, while that of 54 Fe was derived from the measurements. 23,24) The total cross section of elemental iron was evaluated on the basis of the experimental data [25][26][27] in the energy region from 700 keV to 20 MeV.…”

Section: Evaluation 1 Light and Medium-heavy Nuclide Datamentioning

confidence: 99%

“…107) The calcula- Table 2 Nuclides for which covariances are given in JENDL-3.3 1 H, 10 B, 11 B, 16 O, 23 Na, 48 Ti, V, 52 Cr, 55 Mn, 56 Fe, 59 Co, 58 Ni, 60 Ni, 90 Fig. 12 that the JENDL-3.3 calculations are in good agreement with the measurements at highly enriched U fuel cores of STACY, TRACY and JRR-4, which is due to the modifications of the fission neutron spectrum and thermal fission cross sections of 235 U.…”

Section: Criticalitiesmentioning

confidence: 99%

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Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3

Shibata

Kawano

Nakagawa

et al. 2002

Journal of Nuclear Science and Technology

447

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Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were solved by the new library: overestimation of criticality values for thermal fission reactors was improved by the modifications of fission cross sections and fission neutron spectra for 235 U; incorrect energy distributions of secondary neutrons from important heavy nuclides were replaced with statistical model calculations; the inconsistency between elemental and isotopic evaluations was removed for medium-heavy nuclides. Moreover, covariance data were provided for 20 nuclides. The reliability of JENDL-3.3 was investigated by the benchmark analyses on reactor and shielding performances. The results of the analyses indicate that JENDL-3.3 predicts various reactor and shielding characteristics better than JENDL-3.2.

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Section: Evaluation 1 Light and Medium-heavy Nuclide Datamentioning

confidence: 99%

Section: Evaluation 1 Light and Medium-heavy Nuclide Datamentioning

confidence: 99%

Section: Criticalitiesmentioning

confidence: 99%

See 1 more Smart Citation

Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3

Shibata

Kawano

Nakagawa

et al. 2002

Journal of Nuclear Science and Technology

447

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“…These models were rigorously used in the evaluation for JENDL-3.2. A multi-step statistical model code EGNASH (45), which is a modified version of GNASH (46), was used for threshold-reaction cross sections and a single-step code CASTHY(47) for inelastic scattering and capture cross sections. The procedures for EGNASH-and CASTHY-KALMAN calculation were strengthened by adding some newly written interfacing programs to the original ones that had been established by K a~a n o ( 4~) .…”

Section: Covariances Estimated From Nuclear Reaction Model Calculationmentioning

confidence: 99%

Evaluation of Covariance Data for Chromium, Iron and Nickel Contained in JENDL-3.2.

Shibata²

1998

J. Nucl. Sci. Technol.

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An evaluation has been made for the covariances of neutron cross sections of 52Cr, 56Fe, 58Ni and 60Ni contained in JENDL-3.2. Reactions considered were the threshold reactions such as (n, 2n), (n, na), (n, n p ) , (n, p ) , (n, d ) , (n, t ) and (n, a ) , the radiative capture reaction above the resonance region, and the inelastic scattering to discrete and continuum levels. Evaluation guidelines and procedures were established during the work.A generalized least-squares fitting code GMA was used in estimating covariances for reactions of which JENDL-3.2 cross sections had been evaluated by taking account of many measured data. For cross sections that had been evaluated by nuclear reaction model calculations, the KALMAN code, which yields covariances of cross sections and of associated model parameters on the basis of the Bayesian statistics, was used in conjunction with reaction model codes EGNASH and CASTHY.The evaluated uncertainties of a few percent to 30% in the cross sections look reasonable, and the correlation matrices show understandable trends. Even though there is no strict way to confirm the validity of the evaluated covariances, tools and procedures adopted in the present work are appropriate for producing covariance files based on JENDL-3.2. The covariances obtained will be compiled into JENDL in the near future. Meanwhile, new sets of optical model and level density parameters were proposed as one of byproducts obtained from the KALMAN calculations.

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“…These models were embodied primarily in three computer programs: COMNUC, 1 * GNASH, 5 and SCAT. 6 The principal results of this work are summarized in Figs.…”

Section: Introductionmentioning

confidence: 99%

Average neutronic properties of prompt fission products

Foster¹,

Arthur²

1982

Self Cite

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We describe here calculations of the average neutronic properties of the ensemble of fission products produced by fast-neutron fission of 235 U and 239 Pu, where the properties are determined before the first beta decay of any of the fragments. For each case we approximate the ensemble by a weighted average over 10 selected nuclides, whose proper¬ ties we calculate using nuclear-model parameters deduced from the systematic properties of other isotopes of the same elements as the fission fragments. The calculations were performed primarily with the COMNUC and GNASH statisticalmodel codes. The results, available in ENDF/B format, in¬ clude cross sections, angular distributions of neutrons, and spectra of neutrons and photons, for incident-neutron ener¬ gies between 10 eV and 20 MeV. Over most of this energy range, we find that the capture cross section of Pu fission frag¬ ments is systematically a factor of two to five greater than for % fission fragments.

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GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra. [In FORTRAN for CDC 7600]

Abstract: CCpvtc j- ,-3-. ur. ! d ;t°.\ c ' r \-+-^-^t^a"ij.aW,. I o s^sZVa I a m o s scientific laboratory of the University of California LOS ALAMOS, NEW MEXICO 87545 An Alfirmalive Action/Equal Opporiunily Employer UNITED STATES

Cited by 53 publications

References 1 publication

Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3

Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3

Evaluation of Covariance Data for Chromium, Iron and Nickel Contained in JENDL-3.2.

Average neutronic properties of prompt fission products

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