Radiation damage and nuclear heating studies in selected functional materials during the JET DT campaign

Lengar, Igor; Čufar, Aljaž; Conroy, S.; Batistoni, P.; Popovichev, S.; Snoj, Luka; Syme, B.; Vila, R.; Stankūnas, Gediminas

doi:10.1016/j.fusengdes.2016.01.033

Cited by 14 publications

(5 citation statements)

References 10 publications

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“…Monte Carlo (MC) transport simulations are an indispensable tool in these neutronics studies, aiding the effort of the design and installation of neutron diagnostics systems [10], as well as analyses of detector response [11][12][13], evaluations of structural embrittlement [14] and the heating of the tokamak's walls and other vital components [15]. The MC method is built on three major pillars, namely on the information about the complex geometry of a tokamak, physics of neutron interaction with the structural materials, and knowledge about the neutron source.…”

Section: Introductionmentioning

confidence: 99%

Multiphysics approach to plasma neutron source modelling at the JET tokamak

et al. 2019

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A novel multiphysics methodology for the computation of realistic plasma neutron sources has been developed. The method is based on state-of-the-art plasma transport and neutron spectrum calculations, coupled with a Monte Carlo neutron transport code, bridging the gap between plasma physics and neutronics. In the paper two JET neutronics tokamak models are used to demonstrate the application of the developed plasma neutron sources and validate them. Diagnostic data for the record JET D discharge 92436 are used as input for the TRANSP code, modelling neutron emission in two external plasma heating scenarios, namely using only neutral beam injection and a combination of the latter and ion cyclotron resonance heating. Neutron spectra, based on plasma transport results, are computed using the DRESS code. The developed PLANET code package is employed to generate plasma neutron source descriptions and couple them with the MCNP code. The effects of using the developed sources in neutron transport calculations on the response of JET neutron diagnostic systems is studied and compared to the results obtained with a generic plasma neutron source. It is shown that, although there are significant differences in the emissivity profiles, spectra shape and anisotropy between the neutron sources, the integral response of the time-resolved ex-vessel neutron detectors is largely insensitive to source changes, with major relative deviations of up to several percent. However it is calculated that, due to the broadening of neutron spectra as a consequence of external plasma heating, larger differences may occur in activation of materials which have threshold reactions located at DD neutron peak energies. The PLANET plasma neutron source computational methodology is demonstrated to be suitable for detailed neutron source effect studies on JET during DT experiments and can be applied to ITER analyses.

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Section: Introductionmentioning

confidence: 99%

Multiphysics approach to plasma neutron source modelling at the JET tokamak

et al. 2019

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“…The gamma-ray transport calculations, performed in the third step of the developed source methodology, were performed with MCNP (version 6.2 [22]) and ENDF/B-VIII.0 nuclear data library for gamma-ray transport [32]. The model used for the validation is an already existing MCNP model of the tokamak JET, which has been used for JET neutronics studies, aiding in the design and installation of neutron diagnostic systems [33], analysis of neutron detector response and their calibration [34,35], the heating of the tokamak's walls and other vital components [36] and neutron streaming through JET structure and torus hall penetrations [37,38]. The model has thus been used for neutron transport calculations inside and outside the tokamak's vacuum vessel as well as in the tokamak's torus hall penetrations.…”

Section: Plasma Gamma-ray Source Modellingmentioning

confidence: 99%

Validation of realistic Monte Carlo plasma gamma-ray source on JET discharges

Žohar¹,

Nocente²,

Kos³

et al. 2022

Nucl. Fusion

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A novel modelling methodology has been developed for the creation of a realistic plasma gamma-ray source for Monte Carlo transport simulations in the tokamak JET. The methodology couples the TRANSP code for plasma transport calculations with the MCNP Monte Carlo particle transport code, thus connecting plasma physics with gamma-ray transport. This paper presents the validation of the developed source methodology by comparing calculated gamma-ray spectra with measurements performed at JET. The validation focuses on gamma-ray spectra measured by the tangential gamma-ray spectrometer during two JET three ion RF scenario discharges, performed in the JET 2019 Deuterium experimental campaign. For validation the calculated plasma gamma-ray spectrum was combined with the neutron induced prompt gamma-ray background, originating in the vacuum vessel, and scaled to absolute values calculating the total number of plasma gamma-ray and neutron emitting reactions. The comparison between calculated and measured gamma-ray spectra shows good agreement with the shape of the calculated gamma-ray spectra matching that of measurements for both studied discharges. Moreover, the calculated absolute values of the gamma-ray spectra were of the same order of magnitude at the position of the gamma-ray detector located at the end of a long line-of-sight in a biological shield. The comparison has validated the developed plasma gamma-ray source methodology for MCNP photon transport calculations at JET. The validation provides a basis for the developed plasma gamma-ray source to be used as a support for the development of future tokamaks such as DEMO.

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“…The module consists of tungsten armor, FW, caps, U‐shaped breeder unit (BU), neutron multiplication zone, manifolds, and the associated gas inlet and outlet pipelines, as illustrated in Figure . Because the neutron‐induced radioactivity of the fusion components depends greatly on the materials utilized in the reactor, the reduced active ferritic/martensitic steel was applied as the structural material. Li‐ceramics Li 4 SiO 4 with 90% 6 Li enrichment and the beryllium pebbles were selected as tritium breeder material and the neutron multiplier material for higher tritium breeding ratio.…”

Section: Blanket Design Descriptionmentioning

confidence: 99%

Design and thermal-hydraulic evaluation of helium-cooled ceramic breeder blanket for China fusion engineering test reactor

Lei

Guo

et al. 2017

Int J Energy Res

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Summary China Fusion Engineering Test Reactor, a fusion tokamak device, is proposed to provide complementary technology and experience for ITER and the future fusion power plant. A helium‐cooled ceramic breeder blanket concept is adopted as the candidate tritium breeding blanket for China Fusion Engineering Test Reactor. Detailed design of the blanket structure located at the outboard equatorial plane is presented. The coolant flow characters in the blanket were calculated by the theoretical method and the finite element method. The comparison of the calculated results was done, and it has a good agreement between theoretical results and simulation results. The results show that the pressure drop is 0.13 MPa and the total temperature rise is 194.6°C.

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Radiation damage and nuclear heating studies in selected functional materials during the JET DT campaign

Cited by 14 publications

References 10 publications

Multiphysics approach to plasma neutron source modelling at the JET tokamak

Multiphysics approach to plasma neutron source modelling at the JET tokamak

Validation of realistic Monte Carlo plasma gamma-ray source on JET discharges

Design and thermal-hydraulic evaluation of helium-cooled ceramic breeder blanket for China fusion engineering test reactor

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