An updated core design for the multi-purpose irradiation facility MYRRHA

Eynde, Gert Van den; Malambu, E.; Stankovskiy, Alexey; Fernandez, Rafaeel; Baeten, P.

doi:10.1080/00223131.2015.1026860

Cited by 33 publications

(19 citation statements)

References 6 publications

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“…4 along its height. It has to be noted that in the reference model the axial position of the beam window has been adjusted to center the axial power release at core mid-plane, the shift of the neutron yield being ∼2 cm upwards [12]. The plot of model average demonstrate the same shift indicating that the models have certain degree of correlation.…”

Section: Reference Beammentioning

confidence: 78%

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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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Section: Reference Beammentioning

confidence: 78%

“…A detailed reference 3D model of current MYRRHA design version 1.6 [12] has been used for particle transport calculations. Vertical and horizontal cross-cuts of the model are shown in Fig.…”

Section: The Modelmentioning

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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“…In this paper, a target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) [4] lead-bismuth cooled fast reactor is performed with the JEFF-3.3 [5] and ENDF/B-VIII.0 [6] state-of-the-art nuclear data libraries and the SUMMON system [7]. Uncertainty quantification (UQ) analyses have already been carried out for previous MYRRHA designs [8][9][10] or with older nuclear data evaluations [11]; nevertheless, since JEFF-3.3 and ENDF/B.VIII.0 were released, an updated analysis was required in order to take into account the changes in the new nuclear data libraries and their impact in the criticality safety coefficients.…”

Section: Introductionmentioning

confidence: 99%

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

Romojaro

García-Herranz

2019

EPJ Web Conf.

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A target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) lead-bismuth cooled fast reactor has been performed with JEFF-3.3 and ENDF/B-VIII.0 state-of-the-art nuclear data libraries and the SUMMON system. Uncertainties in keff due to uncertainties in nuclear data have been assessed against the target accuracies provided by SG-26 of the WPEC of OECD/NEA in 2008 for LFR. Results show that keff target accuracy is still exceeded by more than a factor of two using the latest nuclear data evaluations released in 2018. Consequently, nuclear data assimilation has been carried out using criticality experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of MYRRHA. The results from this work show that the level of accuracy needed in nuclear data cannot be obtained using only differential experiments, but the combination of experimental covariance data and integral experiments together with Generalised Least Squares technique can provide adjusted nuclear data capable of predicting reactor properties with lower uncertainty and consistent with differential data.

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“…During the last two years of this project, four critical VENUS-F reactor cores were investigated. These zero power cores represented the features of the lead cooled fast core designs of the ADS MYRRHA [2] and LFR ALFRED [3]. The composition of the fuel assemblies (FA) for these cores consisted of enriched metallic uranium, lead and Al 2 O 3 rodlets to simulate the oxide fuel of MYRRHA and ALFRED.…”

Section: Introductionmentioning

confidence: 99%

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

et al. 2017

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Abstract. During the GUINEVERE FP6 European project (2006)(2007)(2008)(2009)(2010)(2011), the zero-power VENUS watermoderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al 2 O 3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

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An updated core design for the multi-purpose irradiation facility MYRRHA

Cited by 33 publications

References 6 publications

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

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

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

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