Concept of the divertor of a fusion neutron source based on a spherical tokamak

Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A.S.; Petrov, V.; Golikov, A.A.; Голубева, А. В.; Goncharov, P. R.; Gryaznevich, M.; Kirnev, G.; Klishchenko, A.V.; Lukyanov, V.; Спицын, А. В.; Sychugov, D. Yu.; Shpansky, Yu.S.

doi:10.1134/s1063780x12060116

Cited by 9 publications

(11 citation statements)

References 35 publications

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“…The idea how to involve the SOL and divertor plasma description into the integrated modelling is described in [10]. The edge and divertor zone are treated with SHM core and two-point edge plasma code [9], and 2D SOL and divertor plasma B2SOLPS5.2 code [8]. Preliminary results of SOL and divertor plasma modelling matched with main plasma parameters are obtained.…”

Section: Sol Plasma Modellingmentioning

confidence: 99%

“…The parameters of a steady state regime and current ramp up stage are adjusted in the integrated modelling using the following codes: ASTRA transport code [2], NUBEAM Monte Carlo code for NBI [3] incorporated into the ASTRA code, DINA free boundary equilibrium and evolution code [4], SPIDER free boundary equilibrium and equilibrium reconstruction code [5], KINX ideal MHD stability code [6], TOKSTAB rigid shift vertical stability code [7], edge and divertor plasma B2SOLPS5.2 code [8] and semi-analytic hybrid modelling code (SHM) of selfconsistent description of the core, edge and divertor plasmas based on the experimental scaling laws [9,10].…”

Section: Introductionmentioning

confidence: 99%

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Integrated modelling of DEMO-FNS current ramp-up scenario and steady-state regime

et al. 2015

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An approach to the integrated modelling of plasma regimes in the projected neutron source DEMO-FNS based on different codes is developed. The integrated modelling allows to eliminate uncertainties in external parameters for such tasks as plasma current ramp up, steady-state plasma consistency, plasma stability and heat load onto the wall and divertor plates. The following codes are employed for the integrated modelling. ASTRA transport code is used for adjustment of the steady-state regime parameters, NUBEAM Monte Carlo code for NBI incorporated into the ASTRA code, DINA free boundary equilibrium and evolution code, SPIDER free boundary equilibrium and equilibrium reconstruction code, KINX ideal MHD stability code, TOKAMEQ free boundary equilibrium code, edge and divertor plasma B2SOLPS5.2 code and Semi-analytic Hybrid Model (SHM) code for self-consistent description of the core, edge and divertor plasmas based on the experimental scaling laws. The consistent steady state regime for the DEMO-FNS plasma and the plasma current ramp up scenario are developed using the integrated modelling approach. Passive copper coils are considered for suppression of the instability to vertical displacement.

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Section: Sol Plasma Modellingmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Integrated modelling of DEMO-FNS current ramp-up scenario and steady-state regime

et al. 2015

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“…Такой подход может быть реализован с использованием пылевых технологий с инжекцией литиевой пыли [38]. Именно он закладывается в концепцию дивертора и первой стенки ТИН-СТ [39].…”

Section: обсуждение результатов моделированияunclassified

Advanced Model for Analysis of Plasma Parameters in a Tokamak With Intense Fusion

Golikov¹,

Kuteev²

2012

PAST-TF

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Усовершенствована «0»-мерная аналитическая модель расчёта параметров стационарного плазменного разряда в токамаке. В модели используются современные скейлинги для удержания энергии плазмы, а также для бутстреп-тока и генерации тока пучками нейтралов. С использованием модели оценены значения плазменного тока, достижимого в стационарных условиях работы токамака. Получены зависимости основных параметров плазмы от характеристик внешних воздействий и рабочих режимов, в том числе от вкладываемой мощности дополнительного нагрева и энергии частиц инжектируемого пучка. В рабочей области нейтронного источника на основе сферического токамака ТИН-СТ (большой радиус R = 0,5 м, малый радиус а = 0,3 м, магнитное поле 1,5 Тл) определены параметры стационарного режима и соответствующие значения мощности реакции синтеза в системе «пучокплазма». Данная модель позволяет оптимизировать параметры плазмы и системы нагрева с целью получения максимальной величины выходной мощности синтеза и интенсивности нейтронного источника. Определена возможная рабочая область для установки ТИН-СТ с мощностью дополнительного нагрева до 10 МВт пучками дейтерия с энергией до 200 кэВ, обеспечивающими достижение значения тока до 1,5 МА при Н-факторе в скейлинге IPB98(y, 2) = 1,5. Показано, что мощность нейтронного потока зависит от параметров инжекции и удержания плазмы и достигает 2 МВт при мощности нагрева 8 МВт. В соответствующих режимах значение нормализованного параметра бета ( N =  t /I N) не превышает 6 при доле энергии, запасенной в быстром компоненте, более 50%.

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Section: Introductionmentioning

confidence: 99%

“…The concept of the divertor and the edge plasma maintenance technology of the DEMO-FNS [3] develops the corresponding concept of the compact neutron source FNS-ST (Fusion Neutron Source based on Spherical Tokamak, see [4]). A variety of modern divertor designs were considered in [4]. The Double Null divertor configuration with long outer divertor legs [5] (≈R/2, R being major plasma radius) and V-shaped corners [6] was accepted.…”

Section: Introductionmentioning

confidence: 99%

Conceptual design of divertor and first wall for DEMO-FNS

et al. 2015

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Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m −2 . Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m −2 and 600-650 °С at 10.5 MW m −2 . The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

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Concept of the divertor of a fusion neutron source based on a spherical tokamak

Cited by 9 publications

References 35 publications

Integrated modelling of DEMO-FNS current ramp-up scenario and steady-state regime

Integrated modelling of DEMO-FNS current ramp-up scenario and steady-state regime

Advanced Model for Analysis of Plasma Parameters in a Tokamak With Intense Fusion

Conceptual design of divertor and first wall for DEMO-FNS

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