ENDF70: A Continuous-Energy MCNP Neutron Data Library Based on ENDF/B-VII.0

Trellue, Holly R.; Little, Robert C.; White, Morgan Curtis; MacFarlane, R.E.; Kahler, Albert C.

doi:10.13182/nt09-2

Cited by 17 publications

(6 citation statements)

References 3 publications

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“…We observed that for response positions B1, B2 and C1 -C4 in the PWR GEN III model, a non-negligible contributing channel was through the lower plenum (see Figs. 9,12,14). This can also be seen from Figs.…”

Section: Responses Below the Coresupporting

confidence: 72%

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On integrating Monte Carlo calculations in and around near-critical configurations – II. Pressurized water reactors

Brovchenko

Burn

Camprini

2023

Annals of Nuclear Energy

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Section: Responses Below the Coresupporting

confidence: 72%

“…As for the PWR GEN II model, the neutron cross-section data were based, where possible, on ENDF/B-VII.1 [12]. The only difference was that thermal neutron S(,) data for hydrogen in light water were based on ENDF/B-VII Release 0 [14] (rather than Release 1 for PWR GEN II [13]).…”

Section: Pwr Gen III With Thick Steel Reflectormentioning

confidence: 99%

On integrating Monte Carlo calculations in and around near-critical configurations – II. Pressurized water reactors

Brovchenko

Burn

Camprini

2023

Annals of Nuclear Energy

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“…We provided the element-wise base composition and microscopic absorption cross sections for each element in Table S1 in the Supporting Information. In our MCNP6 simulations, we specified the isotropic composition and used the ENDF70 Continuous-Energy MCNP Neutron Data Library (Trellue et al, 2009), which computes the flux-dependent microscopic cross sections. The microscopic cross sections in Table S1 are for 2.2-km/s neutrons at 293 K (Sears, 1992) and were used only for reporting the value of the macroscopic absorption cross-section parameter throughout our analysis (i.e., it was not used directly in the simulations).…”

Section: Modeled Data Setmentioning

confidence: 99%

Analysis of Active Neutron Measurements From the Mars Science Laboratory Dynamic Albedo of Neutrons Instrument: Intrinsic Variability, Outliers, and Implications for Future Investigations

et al. 2020

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The Dynamic Albedo of Neutrons (DAN) instrument onboard Mars Science Laboratory uses neutron die-away, an active nuclear spectroscopy technique, to measure the abundance and depth distribution of hydrogen and neutron-absorbing elements (e.g., Fe and Cl) in the top ∼0.5 m of the Mars subsurface. We examined the intrinsic variability in neutron die-away data using simulated DAN measurements with a range of compositions relevant to equatorial and high-latitude environments on Mars. Our analysis shows that the total neutron counts and timing of neutron arrival explain most variability in thermal neutron die-away curves for a DAN-like instrument configuration over homogeneous subsurfaces. We analyzed the variability in thermal neutron die-away across a variety of elemental compositions that might be observed by DAN or future instruments. We found that when H concentration is low (e.g., in equatorial regions including Gale crater), neutron die-away is most sensitive to variations in H. Conversely, when H concentration is high (e.g., in poleward regions of Mars or icy bodies), neutron die-away is most sensitive to variations in neutron absorbers. We estimated that the H and neutron absorber concentrations in DAN measurements acquired Sols 1-2080 in Gale crater and found that most measurements had low H and moderate absorber content. We presented an outlier analysis of these measurements to identify targets with high potential scientific impact and found that most outliers along the traverse corresponded with abnormally low or high neutron absorber content measurements, some of which may be associated with high-silica fracture-associated halos and felsic igneous material. Plain Language Summary The Dynamic Albedo of Neutrons (DAN) instrument on the MarsScience Laboratory (MSL) Curiosity rover uses active nuclear spectroscopy to measure the abundance and depth distribution of hydrogen, chlorine, and iron in the Martian subsurface. We presented the first comprehensive study of the ways in which active neutron spectroscopy measurements and the sensitivity vary with changes in elemental composition and found that this technique is most sensitive to changes in hydrogen when the abundances of these elements in the subsurface are relatively low, as in Gale Crater. In contrast, we found that when hydrogen abundance is relatively high (e.g., in poleward regions of Mars or on icy bodies), this technique is most sensitive to changes in iron and chlorine (and other neutron-absorbing elements). In addition, we performed an outlier analysis of all DAN measurements acquired between Sols 1 and 2080 and found 46 outliers that correspond to rocks and soils with anomalous iron and chlorine content that may indicate targets of particular scientific interest. We plan to use the presented outlier detection method to quickly analyze new measurements as they are acquired by DAN on MSL and identify those with potentially high scientific value before the rover drives away from each site.

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“…Simulations of neutron flux in aluminium samples and simulations of proton and deuteron flux in lead samples were performed in Monte Carlo simulation code MCNPX version 2.7 [8]. Neutron data library used for the simulations was ENDF70 [9] and physics models were INCL4.2 [10] + ABLA-KHSv3p [11,12]. Cross sections of above mentioned reactions were calculated by deterministic code TALYS version 1.6 [13].…”

Section: Tab 1: Beam Characteristics Of Thementioning

confidence: 99%

Determination And Monte Carlo Simulations Of Neutron Flux Inside Spallation Target Quinta

Tichý¹,

Adam²,

Baldin³

et al. 2017

Proceedings of the 26th International Nuclear Physics Conference — PoS(INPC2016)

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This paper deals with experiments performed with natural uranium spallation target QUINTA irradiated by 4 GeV and 8 GeV deuteron beam from Nuclotron accelerator and by 660 MeV proton beam from Phasotron accelerator. Reaction rates for 24 Na, 57 Co, 58 Co, 205 Bi and 206 Bi production in Al, Co and Pb threshold radioactivation foils were determined. The data served for calculation of the experimental neutron flux inside the QUINTA target for the Phasotron experiment. Simulations in Monte Carlo code MCNPX 2.7 were performed and compared to the experimental data. In the current state of research, the data evaluation and simulations are preliminary.

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ENDF70: A Continuous-Energy MCNP Neutron Data Library Based on ENDF/B-VII.0

Abstract: Following the release of ENDF/B-VII.0 evaluations, an ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced at

Cited by 17 publications

References 3 publications

On integrating Monte Carlo calculations in and around near-critical configurations – II. Pressurized water reactors

On integrating Monte Carlo calculations in and around near-critical configurations – II. Pressurized water reactors

Analysis of Active Neutron Measurements From the Mars Science Laboratory Dynamic Albedo of Neutrons Instrument: Intrinsic Variability, Outliers, and Implications for Future Investigations

Determination And Monte Carlo Simulations Of Neutron Flux Inside Spallation Target Quinta

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