• Nuclear Engineering and Technology
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• 제54권3호
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• pp.817-825
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• 2022

A recent benchmarking effort, under the auspices of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA), has been made to evaluate the current state of modeling and simulation tools available to model fluoride salt-cooled high temperature reactors (FHRs). The FHR benchmarking effort considered in this work consists of several cases evaluating the neutronic parameters of a 2D prismatic FHR fuel assembly model using the participants' choice of simulation tools. Benchmark participants blindly submitted results for comparison with overall good agreement, except for some which significantly differed on cases utilizing a molybdenum-bearing control rod. Participants utilizing more recently updated explicit isotopic cross sections had consistent results, whereas those using elemental molybdenum cross sections observed reactivity differences on the order of thousands of pcm relative to their peers. Through a series of supporting tests, the authors attribute the differences as being nuclear data driven from using older legacy elemental molybdenum cross sections. Quantitative analysis is conducted on the control rod to identify spectral, reaction rate, and cross section phenomena responsible for the observed differences. Results confirm the observed differences are attributable to the use of elemental cross sections which overestimate the reaction rates in strong resonance channels.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2002년도 추계 학술발표회 논문집
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• pp.71-79
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• 2002

이중구조 가연성독봉(Duplex BP)의 성능을 평가하기 위해 한국표준형발전소 24개월 주기를 기준으로 16개 Gadolinia 독봉이 장전된 핵연료집합체에 대해 핵적 평가를 수행하였다. 16개 Gd 독봉이 장전된 핵연료집합체와 동일한 반응도 억제가를 갖는 Duplex 독봉집합체를 설계하기 위해 내심에 Natural U-12wt%Gd$_2$O$_3$, 외심에는 4.95wt%$UO_2$-2w/oEr$_2$O$_3$을 넣어 이중 성형한 24개의 이중구조 가연성독봉이 장전된 핵연료집합체를 설계하였다. 또한 같은 방법으로 140개의 Erbia 독봉이 장전된 등가핵연료집합체를 설계하였다. 핵설계 특성평가를 위해 연소도에 따른 무한증배계수, 반응도억제가, 첨두봉출력 그리고 냉각재 온도재수에 대한 변화에 대해서 비교하였다. Duplex 독봉은 Gadolinia 독봉에 비해 k-inf의 2차 첨두현상을 완화시켜 반응도 제어면에서 유리한 것으로 나타났다. 그러나, 다량의 Erbia 독봉을 전체적으로 골고루 장전한 핵연료집합체보다는 Duplex BP를 장전한 핵연료집합체가 노심내 반응도 제어면에서 유리하지 못한 것으로 나타났다.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.317-324
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• 1987

본 논문은 가압 경수형 원자로의 제어봉 이탈사고와 같이 공간 의존적 과도특성 해석에 필히 요구되는 가상적 사고 분석을 위한 핵적 동특성 코드의 개발을 위한 것이다. 본 논문에서는 1.5군 중성자 화산 방정식에 의거한 수정형 Borresen 모형을 핵적 동특성 모델로 잡고 이를 공간의존적 과도특성해석에 응용할 수 있도록 수식화 하여 고리 1호기 초기 노심의 가상적인 제어봉 이탈 사고해석에 응용했다. 본 사고 해석에 채택한 수정형 Borresen 모형에 대한 계산 정밀도의 검증을 위해 출력 분포 및 제어봉가등 계산결과를 고리 1호기 초기 노심의 노물리 실험자료와 비교했고 공간의존적 사고해석에 있어서 중시되는 핵적 동특성 방정식의 계산 효율성을 검토했다. 그리고 이 결과를 토대로 수정형 Borresen 모형이 제어봉 이탈사고, 증기관 파탄사고 등과 같은 공간의존적 사고해석에 유용하게 이용될 수 있다는 것을 보였다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.