• Nuclear industry
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• v.6 no.10 s.44
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• pp.25-30
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• 1986

지난 4월 26일 발생한 소련 체르노빌 원자력발전소 사고로 전세계의 관심을 집중시켰던 소련의 흑연채널형로(RBMK)에 대하여서는 자유세계에 사고내용과 설비현황이 수시로 보도된 바 있으나, 소련에는 RBMK형과 같이 쌍벽을 이루고 있는 로형의 하나로 소련이 독자적으로 개발하여 상용화하고 있는 가압수형로(VVER)가 있다. 체르노빌원전 사고후 소련의 원자력이용국가위원회 부의장은 '앞으로 소련은 PWR 개발에 주력하겠다'고 발표한 바 있어, 현재 건설 및 계획중인 VVER형이 장차 주동력로가 될 것으로 예상되며, RBMK형에서 VVER형으로 전환될 것으로 생각된다. 최근의 보도에 의하면 소련이 북한에 총설비용량 176만KW의 원자력발전설비를 공급한다고 하여 국내 원자력계에 지대한 관심을 불러일으키고 있는데, 현재까지 소련이 위성국가에 공급한 발전로가 대부분 VVER-440형이며, 북한의 전력설비용량으로 보아 단기용량 44만KW가 기술적으로 최적이고 또한 소련이 기술자립이된 로형이란 점 등으로 소련이 북한에 공급할 원자로는 VVER-440형 4기가 될 가능성이 가장 크다고 추정되므로 이번에 특집으로 VVER형로에 관한 자료와 아울러 소련 및 위성국가들의 원자력개발 실정을 정리 소개한다.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.259-270
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• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1762-1776
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• 2024

A spent nuclear fuel (SNF) analysis module for the Vodo-Vodyanoi Energetichesky Reactor (VVER) was developed and validated in this study. This advancement expands the application area of the existing nodal diffusion code, RAST-V, and reduces the need for additional code during 3D core simulations for SNF analysis, leading to increased efficiency in simulation time. RAST-V uses Lagrange interpolation and a power correction factor derived from the Bateman equation to bypass the re-depletion calculations, which are used to solve the microdepletion chain. This approach improved the efficiency of analysis. To mirror the conditions during the 3D core simulations, the module used history indices related to the moderator temperature, fuel temperature, and boron concentration. The module can predict 1620 isotopes. This paper presents the validation of this isotope inventory prediction and the application of burnup credit. The VVER analysis module was validated using 28 samples discharged from the Novovoronezh-4. Most isotopes were within 10 % of the boundaries of the measurements. This study successfully offers verification results using VVER benchmarks and discusses the application of burnup credit using a VVER-440 cask.

• Nuclear industry
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• v.6 no.10 s.44
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• pp.44-50
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• 1986

소련제 가압수형경수로(VVER)를 도입, 착실한 운전실적을 쌓고 있는 체코슬로바키아에서의 VVER형원전 건설 및 운전경험 등 그 이용 현황을 소개한다.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.814-826
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• 2015

The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid ($TiO_2$/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.1-18
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• 2021

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

• Nuclear industry
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• v.6 no.10 s.44
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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.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3494-3515
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• 2022

This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo-vodyanoi energetichesky reactor) analysis. A VVER analytic solver has been implemented in an in-house nodal diffusion code, RAST-K. The new RAST-K version, RAST-V, uses the triangle-based polynomial expansion nodal method. The RAST-K code provides stand-alone and two-step computation modes for steady-state and transient calculations. An in-house lattice code (STREAM) with updated features for VVER analysis is also utilized in the two-step method for cross-section generation. To assess the calculation capability of the formulated analysis module, various verification and validation studies have been performed with Rostov-II, and X2 multicycles, Novovoronezh-4, and the Atomic Energy Research benchmarks. In comparing the multicycle operation, rod worth, and integrated temperature coefficients, RAST-V is found to agree with measurements with high accuracy which RMS differences of each cycle are within ±47 ppm in multicycle operations, and ±81 pcm of the rod worth of the X2 reactor. Transient calculations were also performed considering two different rod ejection scenarios. The accuracy of RAST-V was observed to be comparable to that of conventional nodal diffusion codes (DYN3D, BIPR8, and PARCS).