Design on Primary Cooling System of Open-Pool Type Research Reactor

Most research reactors are designed as an open-pool type and the reactor is located on the bottom of the open-pool. The reactor in the pool is connected to the primary cooling system, which is designed for adequate cooling of the heat generated from the reactor core. One of the characteristics of an open-pool type research reactor is that the primary coolant after passing through the reactor core and the primary cooling system (PCS) is returned to the reactor pool. Because the primary coolant contains many kinds of radionuclides, the research reactor should be designed to protect the radionuclides from being released outside the pool by a stratified stable water layer, which is formed between a hot water layer and cold water near the reactor and prevents the natural circulation of water in the pool. In this study, additional components such as a discharge header and a working platform inside the pool were developed to help diminish the radiation level to the pool top. To discharge coolant stably inside the reactor pool, a discharge header was installed at the end of the pool inlet pipe. Many holes were made in the discharge header to discharge the coolant slowly and minimize the disturbance of the hot water layer by the flow inside the pool. The working platform was also equipped inside the reactor pool to remove the convective flow near the pool top. The commercially available CFD code, ANSYS CFD-FLEUNT, was used to specifically design the discharge header and working platform for satisfying the requirement of the pool top radiation level. The computations were conducted to analyze the flow and temperature characteristics inside the pool for several geometries using an SST k-ω turbulent model and cell modeling, which were conducted to isolate the root cause of these differences and the given inlet conditions. The discharge header and working platform were designed using the CFD results.

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Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽

10.3139/124.110878 ◽

2018 ◽

Vol 83 (3) ◽

pp. 232-236 ◽

Cited By ~ 1

Author(s):

D. L. Zhang ◽

P. Song ◽

S. Wang ◽

X. Wang ◽

J. Chen ◽

...

Keyword(s):

Fast Reactor ◽

Cooling System ◽

Pool Type ◽

Code Development

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Study on the Heat Transfer in the Water Pool Type Reactor Cavity Cooling System of the Very High Temperature Gas Cooled Reactor

Heat Transfer: Volume 4 ◽

10.1115/ht2005-72589 ◽

2005 ◽

Author(s):

H. K. Cho ◽

D. U. Seo ◽

M. O. Kim ◽

G. C. Park

Keyword(s):

Heat Transfer ◽

Forced Convection ◽

Cooling System ◽

Reactor Vessel ◽

Water Pool ◽

Cooling Pipe ◽

Cavity Cooling ◽

The Heat Transfer Coefficient ◽

Pool Type ◽

High Temperature Gas

In the HTGR (High Temperature Gas Cooled Reactor), the Reactor Cavity Cooling System (RCCS) is equipped to remove the heat transferred from the reactor vessel to the structure of the containment. The function of the RCCS is to dissipate the heat from the reactor cavity during normal operation including shutdown. The system also removes the decay heat during the loss of forced convection (LOFC) accident. A new concept of the water pool type RCCS was proposed at Seoul National University. The system mainly consists of two parts, water pool located between the containment and reactor vessel and five trains of air cooling system installed in the water pool. In normal operations, the heat loss from the reactor vessel is transferred into the water pool via cavity and it is removed by the forced convection of air flowing through the cooling pipes. During the LOFC accident, the after heat is passively removed by the water tank without the forced convection of air and the RCCS water pool is designed to provide sufficient passive cooling capacity of the after heat removal for three days. In the present study, experiments and numerical calculations using CFX5.7 for the water pool and cooling pipe were performed to investigate the heat transfer characteristics and evaluate the heat transfer coefficient model of the MARS-GCR (Multi-dimensional Analysis of Reactor Safety for Gas Cooled Reactor Analysis) which was developed for the safety analysis of the gas cooled reactor. From the results of the experiments and CFX calculations, heat transfer coefficients inside the cooling pipe were calculated and those were used for the assessment for the heat transfer coefficient model of the MARS-GCR.

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OPTIMASI DESAIN TERMOHIDROLIKA TERAS DAN SISTEM PENDINGIN REAKTOR RISET INOVATIF DAYA TINGGI

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2015.17.3.2327 ◽

2015 ◽

Vol 17 (3) ◽

pp. 127 ◽

Cited By ~ 2

Author(s):

Endiah Puji Hastuti ◽

Muhammad Subekti ◽

Sukmanto Dibyo ◽

M. Darwis Isnaini

Keyword(s):

Conceptual Design ◽

High Power ◽

Research Reactor ◽

Cooling System ◽

Flow Boiling ◽

Reactor Core ◽

System Optimization ◽

Hydraulic Calculation ◽

Design System ◽

Steady State Condition

ABSTRAK OPTIMASI DESAIN TERMOHIDROLIKA TERAS DAN SISTEM PENDINGIN REAKTOR RISET INOVATIF DAYA TINGGI. Implementasi reaktor inovasi telah diterapkan pada berbagai reaktor riset baru yang saat ini sedang dibangun. Pada saat ini BATAN sedang merancang desain konseptual reaktor riset daya tinggi yang telah masuk pada tahap optimasi desain. Spesifikasi desain konseptual reaktor riset inovatif adalah reaktor tipe kolam berpendingin air dan reflektor D2O. Teras reaktor memiliki kisi 5x5 dengan 16 bahan bakar dan 4 batang kendali. Teras reaktor berada di dalam tabung berisi D2O yang berfungsi sebagai posisi iradiasi. Daya reaktor 50 MW didesain untuk membangkitkan fluks neutron termal sebesar 5x1014 n/cm2s. Teras reaktor berbentuk kompak dan menggunakan bahan bakar U9Mo-Al dengan tingkat muat uranium 7-9 gU/cm3. Desain termohidrolika yang mencakup pemodelan, perhitungan dan analisis kecukupan pendingin dibuat sinergi dengan desain fisika teras agar keselamatan reaktor terjamin. Makalah ini bertujuan menyampaikan hasil analisis perhitungan termohidrolika teras dan sistem reaktor riset inovatif pada kondisi tunak. Analisis dilakukan menggunakan program perhitungan yang telah tervalidasi, masing-masing adalah Caudvap, PARET-ANL, Fluent dan ChemCad 6.4.1. Hasil perhitungan menunjukkan bahwa pembangkitan panas yang tinggi dapat dipindahkan tanpa menyebabkan pendidihan dengan menerapkan desain teras reaktor bertekanan, di samping itu desain awal komponen utama sistem pembuangan panas yang terintegrasi telah dilakukan, sehingga konseptual desain termohidrolika RRI-50 dapat diselesaikan. Kata kunci : reaktor riset inovatif, Caudvap, PARET-ANL, Fluent, ChemCad 6.4.1. ABSTRACT THERMALHYDRAULIC DESIGN AND COOLING SYSTEM OPTIMIZATION OF THE HIGH POWER INOVATIVE RESEARCH REACTOR. Reactor innovation has been implemented in a variety of new research reactors that currently are being built. At this time BATAN is designing a conceptual design of the high power research reactor which has entered the stage of design optimization. The conceptual design specifications of the innovative research reactor is a pool type reactor, water-cooled and reflected by D2O. The reactor core has a 5 x 5 grid with 16 fuels and 4 control rods, which is inserted into a tube containing D2O as an irradiation position. Reactor power of 50 MW is designed to generate thermal neutron flux of 5x1014 n/cm2s. The compact core reactor is using U9Mo-Al fuel with uranium loading of 7-9 gU/cm3. Thermal hydraulic design includes modeling, calculation and analysis of the adequacy of coolant created synergy with the physical design of reactor safety. This paper aims to deliver the results of thermal hydraulic calculation and system design analysis at steady state condition. The analysis was done using various calculation programs that have been validated, i.e. Caudvap, PARET-ANL, Fluent and ChemCad 6.4.1. The calculation results show that the heat generation can be transfered without causing a two phase flow boiling by applying pressurized reactor core design, while the main components of initial design system with an integrated heat dissipation has been done, to complete the conceptual design of the RRI-50 thermalhydraulics. Keywords : inovative research reactor, Caudvap, PARET-ANL, Fluent, ChemCad 6.4.1.

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Evaluation of technological radiation sources in the pumping station of the primary circulation loop and on the top of the pool type research reactor

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp1801031k ◽

2018 ◽

Vol 33 (1) ◽

pp. 31-46

Author(s):

Stoyan Kadalev

Keyword(s):

Heat Carrier ◽

Decay Time ◽

Research Reactor ◽

Irradiation Time ◽

Nuclear Research ◽

Circulation Loop ◽

Nuclear Facilities ◽

Energy Agency ◽

Radiation Sources ◽

Pool Type

The present paper considers the approach to an assessment of technological radiation sources in the primary water-water reactor circulation loop. In principle, such an evaluation is a multidisciplinary task that covers not only the irradiation of the nuclei, the formation of new isotopes and their decay when they are unstable, but also calculations in the field of hydraulics in order to perform an assessment of the irradiation time and the decay time. A general and a more detailed review of the radiation sources formation in the nuclear facilities and the pool type research reactors with demineralized water as a heat carrier are prepared. The initial isotopic composition of the heat carrier has been adopted according to the Vienna Standard Mean Ocean Water recommended by the International Atomic Energy Agency. The general mathematical model of the processes of nuclei irradiation, the formation of new isotopes and their decay, the assessment of the irradiation time and the decay time is described in details, enabling the repetition of this evaluation to a particular facility. The presented approach is applied in the reconstruction design of the nuclear research reactor IRT-2000, Sofia, Bulgaria.

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