Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

Košťál, Michal; Švadlenková, Marie; Baroň, P.; Rypar, Vojtěch; Milčák, Ján

doi:10.1016/j.anucene.2015.12.028

Cited by 24 publications

(3 citation statements)

References 11 publications

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“…As the difference between the moderator level and the position of the activation foil increases, the underestimation rises. This phenomenon was previously observed and described in different type of experiment with a partially flooded reactor core, see [17] .…”

Section: A Reaction Rate Measurement In Lr-0 Graphite Coresupporting

confidence: 72%

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Peltan,

Vilimova,

Czakoj

et al. 2023

EPJ Web Conf.

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This paper is focused on the development of the experimental environment connected to reactor graphite. Regarding its very good neutronic and mechanical properties, graphite will be very important in some new reactor designs, such as high-temperature or molten salt SMR reactors. These new reactor concepts require a new experimental environment as support for further research. In the laboratories of the Research Centre Řež and at the LR-0 reactor, the new experimentally validated graphite environment was created. This large graphite insertion is the largest graphite mono-block, which is possible to assemble at the LR-0 reactor. Sets of experiments for measuring reaction rates of different activation detectors for neutron field mapping were performed. This approach was used for thermal and epithermal region descriptions. For the fast neutron spectrum evaluation, the stilbene scintillation detector was used. All parameters, such as criticality height of moderator level, neutron spectrum, and other parameters for all experiments, were performed using Monte Carlo neutronic codes Serpent and MCNP. The obtained results were finally compared to the measurement of neutron leakage spectra from the graphite cube and graphite cylinder. These specially developed graphite-shaped neutron fields, reactor insertions, and external cube and cylinder with Cf neutron source can be used in the future for validation of not only materials used in SMR reactors but for arbitrary cross-section verification.

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Section: A Reaction Rate Measurement In Lr-0 Graphite Coresupporting

confidence: 72%

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Peltan,

Vilimova,

Czakoj

et al. 2023

EPJ Web Conf.

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show abstract

“…Reactor criticality is achieved by variation of a light water level only. Characterization of the special core was performed by the experiments dealing with reactivity characterization [6], fission rates distribution [7], and also neutron spectra measurement using a stilbene scintillation detector [8]. Details about SACS measurements in LR-0 can be found in [9], [10], [11].…”

Section: Neutron Sources Descriptionmentioning

confidence: 99%

Spectral averaged cross sections as a probe to a high energy tail of ²³⁵U PFNS

et al. 2023

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The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf neutron source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates which were derived from gamma spectrometry. These gamma spectrometry measurements were performed using a single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as the 252Cf spontaneous fission spectrum is considered as a standard. Furthermore, ratios are independent of cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting it to be harder than in other evaluations.

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“…1). It is a well-defined core with an already characterized reactivity [6], fission rates distribution [7] and also neutron spectra [8] (see Fig. 2).…”

Section: Experimental Methodsmentioning

confidence: 99%

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

et al. 2017

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Abstract. The presented paper aims to compare the calculated and experimental reaction rates of 23 Na(n,2n) 22 Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252 Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination.Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of longterm activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3

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Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

Cited by 24 publications

References 11 publications

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Spectral averaged cross sections as a probe to a high energy tail of ²³⁵U PFNS

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra

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Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

Cited by 24 publications

References 11 publications

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Specially developed LR-0 reactor graphite environment for gen IV reactor support and cross-section measurement

Spectral averaged cross sections as a probe to a high energy tail of 235U PFNS

Measurement of the23Na(n,2n) cross section in235U and252Cf fission neutron spectra

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Product

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Spectral averaged cross sections as a probe to a high energy tail of ²³⁵U PFNS

Measurement of the²³Na(n,2n) cross section in²³⁵U and²⁵²Cf fission neutron spectra