• Title, Summary, Keyword: VVER-1000

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소련의 원자력개발과 VVER형원전 - 가압수형 경수로(VVER)의 개요

  • 한국원자력산업회의
    • Nuclear industry
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    • v.6 no.10
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    • pp.36-43
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    • 1986
  • 소련 Voronezh시 근교에 소련 50주년을 기념하여 Novo-Voronezh원자력발전소가 착공되어 1964년 9월 원전을 개시하였다. 이 원전은 전기출력 21만KW의 경수로로서 감속재와 냉각재로 보통의 물(경수)을 사용하였다. 1호기의 발전개시후 2호기의 설계에 착수, 1호기의 문제점을 피드.백(Feed Back)하여 1969년 12월 2호기가 운전을 시작하였다. 전기출력은 36만 KW로 상승하였고 발전코스트는 1호기보다 $40\%$싸며 신뢰성이 확인되어 화력발전보다 경제성 우위가 인정돼 소련방발전전화성이 정식 로형으로 채택하게 되었다. 그후 설비개량을 거듭하여 전기출력 44만KW(VVER-440)을 표준으로 하였으며, 현재는 출력을 증가시켜 100만KW(VVER-1000)와 같이 소련 경수로(PWR)의 주력설비로 되었다. 다음은 VVER의 주요설비개요이다.

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SEVERE ACCIDENT MANAGEMENT CONCEPT OF THE VVER-1000 AND THE JUSTIFICATION OF CORIUM RETENTION IN A CRUCIBLE-TYPE CORE CATCHER

  • Khabensky, Vladimir Benzianovich;Granovsky, Vladimir Semenovich;Bechta, Sevostian Victorovich;Gusarov, Victor Vlasmirovich
    • Nuclear Engineering and Technology
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    • v.41 no.5
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    • pp.561-574
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    • 2009
  • First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

Neutronics design of VVER-1000 fuel assembly with burnable poison particles

  • Tran, Hoai-Nam;Hoang, Van-Khanh;Liem, Peng Hong;Hoang, Hung T.P.
    • Nuclear Engineering and Technology
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    • v.51 no.7
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    • pp.1729-1737
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    • 2019
  • This paper presents neutronics design of VVER-1000 fuel assembly using burnable poison particles (BPPs) for controlling excess reactivity and pin-wise power distribution. The advantage of using BPPs is that the thermal conductivity of BPP-dispersed fuel pin could be improved. Numerical calculations have been conducted for optimizing the BPP parameters using the MVP code and the JENDL-3.3 data library. The results show that by using $Gd_2O_3$ particles with the diameter of $60{\mu}m$ and the packing fraction of 5%, the burnup reactivity curve and pin-wise power distribution are obtained approximately that of the reference design. To minimize power peaking factor (PPF), total BP amount has been distributed in a larger number of fuel rods. Optimization has been conducted for the number of BPP-dispersed rods, their distribution, BPP diameter and packing fraction. Two models of assembly consisting of 18 BPP-dispersed rods have been selected. The diameter of $300{\mu}m$ and the packing fraction of 3.33% were determined so that the burnup reactivity curve is approximate that of the reference one, while the PPF can be decreased from 1.167 to 1.105 and 1.113, respectively. Application of BPPs for compensating the reduction of soluble boron content to 50% and 0% is also investigated.

Application of a new neutronics/thermal-hydraulics coupled code for steady state analysis of light water reactors

  • Safavi, Amir;Esteki, Mohammad Hossein;Mirvakili, Seyed Mohammad;Arani, Mehdi Khaki
    • Nuclear Engineering and Technology
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    • v.52 no.8
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    • pp.1603-1610
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    • 2020
  • Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.