Impact of the ENDF/B-VIII.0 Library on Advanced Reactor Simulations

Bostelmann, Friederike; Holcomb, Andrew; Marshall, William J.; Rearden, Bradley T.; Sobes, Vladimir

doi:10.13182/t31078

Cited by 5 publications

(11 citation statements)

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“…As applicable to the other discussed reactor concepts, the updates between the ENDF/B-VII.1 and ENDF/B-VIII.0 releases of relevant cross sections of 16 O, 56 Fe, 238 U, and 239 Pu, for example, cause differences in reactivity results. Reactivity differences of up to 200 pcm were observed in previous studies due to updates in 238 U data (Bostelmann et al 2018a, Bostelmann et al 2019.…”

Section: Sodium-cooled Fast Reactormentioning

confidence: 64%

“…In a previous ORNL study, significant differences of a few hundred pcm in multiplication factors for advanced reactor models were noted when comparing results obtained with ENDF/B-VII.1 data to results obtained with ENDF/B-VIII.0 (Bostelmann et al 2018b). Starting with an ENDF/B-VII.1 calculation and replacing the nuclear data of individual nuclides with data from ENDF/B-VIII.0 one nuclide at a time, it was found that updates in the cross sections of 235 U and 238 U were responsible for most observed differences in multiplication factors (Bostelmann et al 2019). Even though these are among the most common nuclides in traditional reactor fuel and their nuclear data are therefore assumed to be very well known, significant changes in the capture, fission, and scattering cross sections were introduced in the latest ENDF/B-VIII.0 release compared to ENDF/B-VII.1.…”

Section: Pebble-bed High-temperature Gas-cooled Reactormentioning

confidence: 99%

“…Due to the large amount of the carrier salt NaCl in the reactor, the scattering and capture reactions of Na and Cl nuclides were also identified as relevant for the reactivity (Bostelmann et al 2018a). Updates of these Na and Cl cross sections as well as for uranium and plutonium cross sections between the ENDF/B-VII.1 and ENDF/B-VIII.0 release can cause reactivity differences of a few hundred pcm (Bostelmann et al 2018a, Bostelmann et al 2019.…”

Section: Molten Chloride Fast Spectrum Reactormentioning

confidence: 99%

“…SFR uncertainty analyses based on the ENDF/B-VIII.0 release are therefore expected to show major differences with respect to output uncertainties and their top contributing nuclide reactions when compared with ENDF/B-VII.1 calculations. The uncertainties of other relevant reactions such as 56 Fe scattering, 239 Pu fission, and 238 U neutron capture were also significantly changed in ENDF/B-VIII.0, so they will contribute to these differences as well (Bostelmann et al 2019).…”

Section: Sodium-cooled Fast Reactormentioning

confidence: 99%

See 3 more Smart Citations

Key Nuclear Data Impacting Reactivity in Advanced Reactors

Bostelmann

Ilas

Wieselquist

2020

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Section: Sodium-cooled Fast Reactormentioning

confidence: 64%

Section: Pebble-bed High-temperature Gas-cooled Reactormentioning

confidence: 99%

Section: Molten Chloride Fast Spectrum Reactormentioning

confidence: 99%

Section: Sodium-cooled Fast Reactormentioning

confidence: 99%

See 2 more Smart Citations

Key Nuclear Data Impacting Reactivity in Advanced Reactors

Bostelmann

Ilas

Wieselquist

2020

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“…Updates of relevant cross sections and uncertainties between the ENDF/B-VII.1 and ENDF/B-VIII.0 releases, including 56 Fe, 238 U, and 239 Pu, can impact reactivity results. For example, reactivity differences of up to 200 pcm were observed in previous studies as a result of updates in 238 U data [9,10].…”

Section: Introductionmentioning

confidence: 73%

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

Bostelmann

Ilas

Wieselquist

2021

JNE

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The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.

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Impact of the ENDF/B-VIII.0 Library on McNP6.2-COMPUTED Ambient and Personal Dose Equivalent Conversion Coefficients

McHale

Millett

Prince

2022

Radiation Protection Dosimetry

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The 2018 release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library initiated several examinations of the impact of the revised nuclear data on Monte Carlo models. This study used version 6.2 of the Monte Carlo N-Particle® code (MCNP6.2) with the ENDF/B-VI and ENDF/B-VIII.0 libraries to compute conversion coefficients for fluence-to-ambient and fluence-to-personal dose equivalent from neutron energy groups described by an International Atomic Energy Agency 53-bin structure and 47 monoenergetic neutron sources listed in International Commission on Radiological Protection (ICRP) 74. The MCNP6.2 models with ENDF/B-VI data were validated against results published in 2005 by Veinot and Hertel. Conversion coefficients computed with MCNP6.2 and ENDF/B-VIII.0 slightly underestimated the ICRP 74 values but were within ICRP-specified tolerances and do not justify revising the ICRP coefficients. Conversion coefficients for personal dose equivalent were computed with MCNP6.2 and ENDF/B-VIII.0 at four angles of incidence greater than those available in ICRP 74. These new coefficients are relevant to radiation protection studies of interest to US defence-focused organisations.

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Impact of the ENDF/B-VIII.0 Library on Advanced Reactor Simulations

Cited by 5 publications

References 0 publications

Key Nuclear Data Impacting Reactivity in Advanced Reactors

Key Nuclear Data Impacting Reactivity in Advanced Reactors

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

Impact of the ENDF/B-VIII.0 Library on McNP6.2-COMPUTED Ambient and Personal Dose Equivalent Conversion Coefficients

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