Thermal hydraulic analysis of a pebble-bed modular high temperature gas-cooled reactor with ATTICA3D and THERMIX codes

Zheng, Yongjun; Lapins, J.; Laurien, Eckart; Shi, Lei; Zhang, Z.

doi:10.1016/j.nucengdes.2012.02.014

Cited by 33 publications

(9 citation statements)

References 7 publications

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“…The ZBS correlation has been applied in thermal hydraulic analysis of packed pebble bed experiments and HTGR [3,6]. It was found that the KuniiSmith correlation and Zehner-Bauer-Schlünder (ZBS) correlation are in good agreement with experimental data at high temperature [21].…”

Section: Existing Correlationsmentioning

confidence: 99%

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Effect of scale on the modeling of radiation heat transfer in packed pebble beds

Hao

Gui

Yang

et al. 2016

International Journal of Heat and Mass Transfer

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Section: Existing Correlationsmentioning

confidence: 99%

“…Moreover, for theoretical and numerical study, the macroscopic models based on the porous-medium assumption of the packed pebble bed have been widely used in the simulation [5] and analysis of transient thermal hydraulics of HTGR [6]. Basically, there are three types of heat transfer modes in the packed pebble beds, i.e.…”

Section: Introductionmentioning

confidence: 99%

Effect of scale on the modeling of radiation heat transfer in packed pebble beds

Hao

Gui

Yang

et al. 2016

International Journal of Heat and Mass Transfer

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“…The results indicated that uniform pressure in the radial direction for a core to fuel element diameter (D/d) ratio > 20 and the heat transfer coefficient increases with increasing temperature and coolant mass flow rate [7]. Zheng et al conducted a research to study Chinese HTR-PM 200 MW reactor characteristics at asymmetric condition [8]. The analysis was done by comparing thermal hydraulic 2-D THERMIX calculation results with the results of 3-D ATTICA3D (Advanced Thermal hydraulics Tool for Invessel & Core Analysis in 3D) reactor system analysis calculation.…”

Section: Introductionmentioning

confidence: 99%

“…The results show that based on symmetric cylindrical power distribution, the two codes give pressure drop and temperature distribution show in good agreement. At asymmetrical condition, in which one set of small absorber spheres (SAS) drops into the side reflectors resulting in an asymmetric distribution of the nuclear power, a three-dimensional calculation for the steady state is implemented to basically show the good 3-D simulation capabilities of ATTICA3D, which will be helpful to the next refined design stage for the HTR-PM project [8]. Nowadays.…”

Section: Introductionmentioning

confidence: 99%

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Investigation on Thermal-Flow Characteristics of HTGR Core Using Thermix-Konvek Module and Vsop'94 Code

Sudarmono

2015

Tri Dasa Mega

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The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR). Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology (PTRKN) has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generation system, which has high energy efficiency, and has high and clean inherent safety level. The geometry and structure of the HTGR200 core are designed to produce the output of helium gas coolant temperature as high as 950 °C to be used for hydrogen production and other industrial processes in co-generative way. The output of very high temperature helium gas will cause thermal stress on the fuel pebble that threats the integrity of fission product confinement. Therefore, it is necessary to perform thermal-flow evaluation to determine the temperature distribution in the graphite and fuel pebble in the HTGR core. The evaluation was carried out by Thermix-Konvek module code that has been already integrated into VSOP'94 code. The HTGR core geometry was done using BIRGIT module code for 2-D model (RZ model) with 5 channels of pebble flow in active core in the radial direction. The evaluation results showed that the highest and lowest temperatures in the reactor core are 999.3 °C and 886.5 °C, while the highest temperature of TRISO UO2 is 1510.20 °C in the position (z= 335.51 cm; r=0 cm). The analysis done based on reactor condition of 120 kg/s of coolant mass flow rate, 7 MPa of pressure and 200 MWth of power. Compared to the temperature distribution resulted between VSOP’94 code and fuel temperature limitation as high as 1600 oC, there is enough safety margin from melting or disintegrating. Keywords: Thermal-Flow, VSOP’94, Thermix-Konvek, HTGR, temperature Kegagalan sistem pembuangan panas pada reaktor berpendingin air jenis PWR, Three Mile Islands dan reaktor BWR Fukushima Daiichi, menyebabkan masyarakat nuklir mulai memikirkan penggunaan reaktor pembangkit daya jenis temperatur tinggi berpendingin gas (HTGR). Bidang Fisika dan Teknologi Reaktor di Pusat Teknologi Reaktor dan Keselamatan Nuklir (PTRKN) mempunyai tugas melaksanakan kegiatan litbang desain konseptual reaktor kogenerasi dengan tingkat daya menengah yang berpendingin gas helium dengan daya 200 MWt. Desain HTGR200K merupakan salah satu sistem pembangkit energi yang memiliki efisiensi energi paling besar, dan tingkat keselamatan inheren yang tinggi dan bersih. Komposisi geometri dan struktur teras didesain agar dapat menghasilkan keluaran pendingin gas helium bertemperatur 950 0C sehingga dapat digunakan untuk produksi hidrogen dan atau unit industri proses lainnya secara kogeneratif. Luaran gas helium bertemperatur sangat tinggi ini akan menimbulkan tegangan termal pada bola bahan bakar yang mengancam integritas sistem pengungkungan produk fisi di dalamnya. Oleh karena itu perlu dilakukan evaluasi karakteristika termal flow untuk menentukan distribusi temperatur bahan bakar bola dan outlet temperatur pendingin gas helium teras HTGR. Hal ini dilakukan dengan menggunakan modul Thermix-Konvek yang terintegrasi dalam program VSOP’94. Geometri teras HTGR dikerjakan dalam modul BIRGIT untuk model teras 2-D (R-Z) dengan 5 kanal aliran pebble dalam teras aktif arah radial. Hasil evaluasi menunjukkan bahwa nilai tertinggi dan terendah temperatur yang terdapat pada teras adalah sebesar 999.3 °C dan 886,5 °C. Demikian pula hasil temperatur tertinggi bahan bakar TRISO dan bahan bakar pebble di dalam teras, yaitu diperoleh sebesar 1510,20°C yang terletak pada lapisan bahan bakar inti UO2, di posisi z= 335.51 cm dan r=0 cm. Analysis di lakukan pada laju massa aliran pendingin, tekanan dan daya masing-masing sebesar 120 kg/s, 7 Mpa dan 200MWth. Hasil perhitungan, jika dibandingkan dengan lisensi pembatas keselamatan terhadap maksimum temperatur bahan bakar pebble menunjukkan bahwa integritas bahan bakar pebble masih aman karena masih lebih rendah dari batas desain yaitu sebesar 1600 oC. Kata kunci: Thermal-Flow, VSOP’94, Thermix-Konvek, HTGR, temperatur

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Numerical Models for Pebble-Bed Heat Transfer

Jiang

Yang

et al. 2020

Multiphase Flow and Heat Transfer in Pebble Bed Reactor Core

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Thermal hydraulic analysis of a pebble-bed modular high temperature gas-cooled reactor with ATTICA3D and THERMIX codes

Cited by 33 publications

References 7 publications

Effect of scale on the modeling of radiation heat transfer in packed pebble beds

Effect of scale on the modeling of radiation heat transfer in packed pebble beds

Investigation on Thermal-Flow Characteristics of HTGR Core Using Thermix-Konvek Module and Vsop'94 Code

Numerical Models for Pebble-Bed Heat Transfer

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