Status of the JEFF Nuclear Data Library

Koning, A. J.; Koning, E.; Dean, John W.; Dupont, E.; Fischer, U.; Forrest, R.A.; Jacqmin, R.; Leeb, H.; Kellett, Mark A.; Mills, Robert W.; Nordborg, C.; Pescarini, M.; Rugama, Y.; Rullhusen, P.

doi:10.3938/jkps.59.1057

Cited by 182 publications

(178 citation statements)

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“…The NaF salt sample and neutron spectra were modeled using these selected nuclear data libraries: ENDF/B-VII.0 [12], JEFF-3.1 [13], JEFF-3.2 [14], JENDL-4 [15], JENDL-3.3 [16], ROSFOND-2010 [17], CENDL-3.1 [18], CIELO [19] and IRDFF [1]. Nuclear data libraries were processed using the NJOY99 code [20].…”

Section: Calculation Methodsmentioning

confidence: 99%

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

et al. 2017

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Abstract. The presented paper aims to compare the calculated and experimental reaction rates of 23 Na(n,2n) 22 Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252 Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination.Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of longterm activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3

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Section: Calculation Methodsmentioning

confidence: 99%

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

et al. 2017

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“…33 Version 2.1.28 of SERPENT is used in combination with the JEFF3.1 nuclear data library. 34 The single-assembly calculations for group-constant generation are run with 750 active cycles of 7.5 × 10 5 source neutrons (50 inactive cycles are discarded to allow the initial fission-source distribution to converge). This choice results in 5.63 × 10 8 active neutron histories.…”

Section: Iva Proceduresmentioning

confidence: 99%

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Gamarino

Dall’Osso

Lathouwers

et al. 2018

Nuclear Science and Engineering

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-Nodal diffusion is currently the preferred neutronics model for industrial reactor core calculations, which use few-group cross-section libraries generated via standard assembly homogenization. The infinite-medium flux-weighted cross sections fail to capture the spectral effects triggered in the core environment by nonreflective boundary conditions at the fuel-assembly edges. This poses a serious limitation to the numerical simulation of current-and next-generation reactor cores, characterized by strong interassembly heterogeneity.Recently, a spectral rehomogenization method has been developed at AREVA NP. This approach consists of an on-the-fly modal synthesis of the spectrum variation between the environmental and infinite-medium conditions. It uses information coming from both the nodal simulation and the lattice transport calculation performed to compute the standard cross sections. The accuracy of the spectral corrections depends on the choice of the basis and weighting functions for the expansion and on the definition of a realistic energy distribution of the neutron leakage. In this paper, we focus on the first aspect. Two tracks are researched: a combination of analytical functions (with a physically justified mode) and a mathematical approach building upon the Proper Orthogonal Decomposition. The method is applied to relevant pressurized-water-reactor benchmark problems. We show that the accuracy of the cross sections is significantly improved at reasonably low computational cost and memory requirement. Several aspects of the methodology are discussed, such as the interplay with space-dependent corrections. We demonstrate that this approach can model not only the spectral interactions between dissimilar neighbor assemblies but also the spectral effects due to different physical conditions (namely, multiplicative properties) in the environment and in the infinite medium.

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“…In order to reconstruct the behavior of the fi ssionable system, fi ve sets of nuclear data libraries representing the best available knowledge about the particle interactions with the matter and natural nuclide properties were linked to the MCB code: -JEFF3.1 continuous neutron transport cross--section libraries for 381 isotopes [10]; -decay schemas for about 2400 nuclides from the Table of Isotopes 8E [11] and dose data for about 738 nuclides from the Euroatom Council Directive 96/29/EUROATOM [12]; -one-group cross-section libraries previously applied in the ORIGEN code containing energy--integrated ratios of the nuclide formation in excited state due to the neutron capture, (n,2n) reaction and nuclear decay [13]; -branching ratios to the ground and metastable state in the function of the incident neutron energy for Am-242m and Am-244m based on the Mann and Schenter's evaluation [14]; -the incident energy depended fi ssion products yields for 36 nuclides based on Wahl model [15]. Figure 6 shows time evolution of lumped uranium containing uranium isotopes present in the initial composition of the nuclear fuel, that is, U-234, U-235, U-236, and U-238.…”

Section: The Mcb Codementioning

confidence: 99%

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

Oettingen

Cetnar

2015

Nukleonika

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In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

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Status of the JEFF Nuclear Data Library

Cited by 182 publications

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Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

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Status of the JEFF Nuclear Data Library

Cited by 182 publications

References 0 publications

Measurement of the23Na(n,2n) cross section in235U and252Cf fission neutron spectra

Measurement of the23Na(n,2n) cross section in235U and252Cf fission neutron spectra

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

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Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra