Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

doi:10.1016/j.nucengdes.2017.02.020

Cited by 10 publications

(8 citation statements)

References 9 publications

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“…The total uncertainty on reactivity is 532 pcm for ARO configuration. It value is close to that calculated in references [16,17]. The greatest contributor to uncertainty is U238 which represents nearly 50% of the uncertainty on the reactivity.…”

Section: Generation Of Four Groups Covariances Matrixsupporting

confidence: 87%

Development and validation of uncertainty neutron transport calculations at an industrial scale

Gaillet

Bonaccorsi

Noguère

et al. 2018

EPJ Nuclear Sci. Technol.

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Evaluating uncertainties on nuclear parameters such as reactivity is a major issue for conception of nuclear reactors. These uncertainties mainly come from the lack of knowledge on nuclear and technological data. Today, the common method used to propagate nuclear data uncertainties is Total Monte Carlo [1] but this method suffers from a long time calculation. Moreover, it requires as many calculations as uncertainties sought. An other method for the propagation of the nuclear data uncertainties consists in using the standard perturbation theory (SPT) to calculate reactivity sensitivity to the desire nuclear data. In such a method, sensitivities are combined with a priori nuclear data covariance matrices such as the COMAC set developed by CEA. The goal of this work is to calculate sensitivites by SPT with the full core diffusion code CRONOS2 for propagation uncertainties at the core level. In this study, COMAC nuclear data uncertainties have been propagated on the BEAVRS benchmark using a two-step APOLLO2/CRONOS2 scheme, where APOLLO2 is the lattice code used to resolve Boltzmann equation within assemblies using a high number of energy groups, and CRONOS2 is the code resolving the 3D full core diffusion equation using only four energy groups. A module implementing the SPT already exists in the APOLLO2 code but computational cost would be too expensive in 3D on the whole core. Consequently, an equivalent procedure has been created in CRONOS2 code to allow full-core uncertainty propagation. The main interest of this procedure is to compute sensitivities on reactivity within a reduced turnaround time for a 3D modeled core, even after fuel depletion. In addition, it allows access to all sensitivites by isotope, reaction and energy group in a single calculation. Reactivity sensitivities calculated by this procedure with four energy groups are compared to reference sensitivities calculated by the iterated fission probability (IFP) method in Monte Carlo code. For the purpose of the tests, dedicated covariance matrix have been created by condensation from 49 to 4 groups of the COMAC matrix. In conclusion, sensitivities calculated by CRONOS2 agree with the sensitivities calculated by the IFP method, which validates the calculation procedure, allowing analysis to be done quickly. In addition, reactivity uncertainty calculated by this method is close to values found for this type of reactor.

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Section: Generation Of Four Groups Covariances Matrixsupporting

confidence: 87%

Development and validation of uncertainty neutron transport calculations at an industrial scale

Gaillet

Bonaccorsi

Noguère

et al. 2018

EPJ Nuclear Sci. Technol.

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“…Otherwise, an overestimation was observed for the outer boundary of the core with the highest difference +7.8%. The obtained deviations are not substantially different from the alternative results available in the literature [8,22,23]. For example in UIUC paper [23], they obtained RMSE equal to 5.38% in comparison to our value 4.89% (see Fig.…”

Section: Resultscontrasting

confidence: 59%

“…It is also possible to observe the lack of small fl ux depressions due to the absence of spacer grid modelling. The axial power distribution calculated with PARCS and reference solution obtained with NECP--VIOLET code ( [8]) is compared in Fig. 5.…”

Section: Resultsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

“…Several research reports on BEAVRS solutions were published for a large variety of simulation methodologies [6][7][8][9][10][11][12][13][14][15][16][17][18][19][20][21][22][23]. The model developed in this work was compared with available benchmark data for the fi rst fuel cycle presented in publications [8,24]. The SCALE-PARCS two-step sequence was tested using BEAVRS data only in a single publicly available conference report, PHYSOR-2018, which describes the University of Illinois at Urbana-Champaign (UIUC) research fi nanced by USNRC [23].…”

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Darnowski

Pawluczyk

2019

Nukleonika

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This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specification contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data file – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS files were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The first fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

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Uncertainty quantification for the first-cycle modeling and simulations of the BEAVRS benchmark problem

Wan

Cao

2019

Annals of Nuclear Energy

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Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

Cited by 10 publications

References 9 publications

Development and validation of uncertainty neutron transport calculations at an industrial scale

Development and validation of uncertainty neutron transport calculations at an industrial scale

Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Uncertainty quantification for the first-cycle modeling and simulations of the BEAVRS benchmark problem

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