Oecd/Nea Benchmark for Uncertainty Analysis in Modeling (Uam) for LWRS – Summary and Discussion of Neutronics Cases (Phase I)

Bratton, Ryan N.; Avramova, Maria; Ivanov, Kostadin

doi:10.5516/net.01.2014.710

Cited by 47 publications

(14 citation statements)

References 2 publications

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“…Results for k eff and two-group macroscopic cross sections at BOC and EOC are shown in Table 2. The calculated uncertainties are comparable to previously compiled results for the pin-cell benchmark [7] and the EOC results clearly show the effect of the larger variances associated with the higher actinides.…”

Section: Pin-cell Calculationssupporting

confidence: 78%

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Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

Wemple¹,

Zwermann

2017

EPJ Web Conf.

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Abstract. Uncertainty quantification has been extensively applied to nuclear criticality analyses for many years and has recently begun to be applied to depletion calculations. However, regulatory bodies worldwide are trending toward requiring such analyses for reactor fuel cycle calculations, which also requires uncertainty propagation for isotopics and nuclear reaction rates. XSUSA is a proven methodology for cross section uncertainty propagation based on random sampling of the nuclear data according to covariance data in multigroup representation; HELIOS2 is a lattice code widely used for commercial and research reactor fuel cycle calculations. This work describes a technique to automatically propagate the nuclear data uncertainties via the XSUSA approach through fuel lattice calculations in HELIOS2. Application of the XSUSA methodology in HELIOS2 presented some unusual challenges because of the highly-processed multi-group cross section data used in commercial lattice codes. Currently, uncertainties based on the SCALE 6.1 covariance data file are being used, but the implementation can be adapted to other covariance data in multi-group structure. Pin-cell and assembly depletion calculations, based on models described in the UAM-LWR Phase I and II benchmarks, are performed and uncertainties in multiplication factor, reaction rates, isotope concentrations, and delayedneutron data are calculated. With this extension, it will be possible for HELIOS2 users to propagate nuclear data uncertainties directly from the microscopic cross sections to subsequent core simulations.

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Section: Pin-cell Calculationssupporting

confidence: 78%

“…The uncertainties are comparable with other compilations of results for this benchmark [7]. With respect to the uncertainties of the delayed neutron fractions, note that the SCALE 6.1 covariance data do not contain uncertainties of the average number of delayed neutrons per fission (ν) separately, but only for the total ν.…”

Section: Depletion Calculationssupporting

confidence: 72%

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

Wemple¹,

Zwermann

2017

EPJ Web Conf.

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“…2 together with the mean value of k inf calculated. Participant-averaged result from exercise I-b benchmark (only available for the final burnup condition in [4]) is presented in the fifth column for comparison purposes. The symbol ∆ is the difference between the value calculated in this study by SAMPLER and the reference value.…”

Section: Resultsmentioning

confidence: 99%

“…This was proposed in scope of the UAM benchmark in order to address uncertainties in nuclear fuel depletion calculation due to the basic nuclear data as well as the impact of processing the nuclear and covariance data. Summary and discussion of neutronic cases performed by the participants is available in [4]. Participantaveraged results for exercise I-b are reproduced here for comparison purposes despite different input covariance cross-sections were used by the participants.…”

Section: Methodsmentioning

confidence: 99%

Uncertainty Propagation for LWR Burnup Benchmark Using Sampling Based Code Scale/Sampler

Campolina

Frýbort

2018

APP

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Abstract.Sampling based method is adopted in many fields of engineering and it is currently used to propagate uncertainties from physical parameters and from nuclear data, to integral indicators of nuclear systems. The total uncertainty associated with a model simulation is of major importance for safety analysis and to guide vendors about acceptable tolerance limits for nuclear installations parts. This work presents some calculations to propagate uncertainties for a nuclear reactor fuel element modeled in SCALE/TRITON, using the sampling tool SCALE/SAMPLER. Results showed that the influence of input uncertainties on k inf is more pronounced in the fresh core other than the depleted core and the contribution from studied manufacturing uncertainties is smaller than the contribution of nuclear data uncertainties.

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“…It has been shown [1][2][3], that the nuclear data uncertainties are most fundamental contributors to the final uncertainty of safetyrelated reactor neutronic parameters. Effective neutron multiplication factor k eff , reactivity effects and neutron kinetic parameters remain fundamental for the sub-critical systems as well.…”

Section: Introductionmentioning

confidence: 99%

Impact of intermediate and high energy nuclear data on the neutronic safety parameters of MYRRHA accelerator driven system

2017

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Abstract. Perturbation of external neutron source can cause significant local power changes transformed into undesired safety-related events in an accelerator driven system. Therefore for the accurate design of MYRRHA sub-critical core it is important to evaluate the uncertainty of power responses caused by the uncertainties in nuclear reaction models describing the particle transport from primary proton energy down to the evaluated nuclear data table range. The calculations with a set of models resulted in quite low uncertainty on the local power caused by significant perturbation of primary neutron yield from proton interactions with lead and bismuth isotopes. The considered accidental event of prescribed proton beam shape loss causes drastic increase in local power but does not practically change the total core thermal power making this effect difficult to detect. In the same time the results demonstrate a correlation between perturbed local power responses in normal operation and misaligned beam conditions indicating that generation of covariance data for proton and neutron induced neutron multiplicities for lead and bismuth isotopes is needed to obtain reliable uncertainties for local power responses.

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Oecd/Nea Benchmark for Uncertainty Analysis in Modeling (Uam) for LWRS – Summary and Discussion of Neutronics Cases (Phase I)

Cited by 47 publications

References 2 publications

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

Uncertainty Propagation for LWR Burnup Benchmark Using Sampling Based Code Scale/Sampler

Impact of intermediate and high energy nuclear data on the neutronic safety parameters of MYRRHA accelerator driven system

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