• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1226-1233
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• 2017

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. The database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.916-921
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• 2024

The control rod worth sensitivity to the neutron detector position in the rod drop experiment is studied by the time-dependent Monte Carlo (TDMC) neutron transport calculations for AGN-201K educational reactor and the Kyoto University Critical Assembly. The TDMC simulations of the rod drop experiments are conducted by the Seoul National University Monte Carlo (MC) code, McCARD, yielding time-dependent neutron densities at detector positions. The detector-position-dependent results of the total control rod worth calculated by the extrapolation, the integral counting, and the inverse methods are compared with the numerical reference using the MC eigenvalue calculations and the experimental results. From these comparisons, it is observed that the total control rod worth can be estimated with a considerable difference depending on the detector position through the rod drop experiment. The proposed TDMC simulation of the rod drop experiment can be applied for searching a better detector position or quantifying a bias for the control rod worth measurement.

• Journal of Advanced Marine Engineering and Technology
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• 제39권6호
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• pp.633-641
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• 2015

선박, 발전소, 석유화학 등의 분야에서 운전되고 있는 많은 프로세스들은 강한 비선형성을 보일 뿐만 아니라 동시에 시변 특성도 가지고 있다. 이런 프로세스에 기존의 고정-파라미터 PID 제어기를 적용하면 성능이 나빠지고 경우에 따라서는 불안정해질 수도 있다. 따라서 본 연구에서는 복잡한 프로세스를 제어하기 위한 비선형 PID 제어기를 제안한다. 제안되는 제어기의 이득은 오차와 오차의 변화율의 비선형 함수로 기술되며, 사용자 파라미터들은 ITAE를 최소로 하는 관점에서 유전알고리즘으로 동조된다. 제안된 방법은 열분해반응 또는 촉매를 이용한 고분자합성에 널리 사용되는 연속 교반탱크반응기를 대상으로 시뮬레이션을 실시하며, 그 유효성을 보이기 위해 다른 두 비선형/적응 제어법과 비교한다.

• 대한기계학회논문집A
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• 제36권6호
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• pp.599-608
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• 2012

일체형 원자로의 안전주입 노즐 이종금속 용접부의 잔류응력에 대한 레이저 피닝의 효과를 고찰하기 위해 상용 프로그램인 ABAQUS를 이용하여 implicit 동적 유한요소 해석을 통해 연구를 수행하였다. implicit 동적 유한요소 해석은 기존의 실험 결과와의 비교에 따르면 레이저 피닝을 통한 잔류응력 이완에 대해 타당하다고 확인된다. 한편 내부 보수용접이 수행된 이종금속 용접부에 대해 해석이 수행되며 그 결과는 축방향 및 원환 잔류응력 모두 내부 보수용접에 기인하여 노즐의 내표면에서 인장임을 나타낸다. 또한 용접 잔류응력 이완에 대한 최대 충격 압력, 압력 지속 시간, 스폿 직경 및 피닝 방향과 같은 여러 변수의 효과를 고찰하기 위해 변수 해석 또한 수행하였다. 결과적으로, 레이저 피닝은 내표면 근처 영역의 잔류응력을 주로 이완시키는 예방정비 효과가 있음을 확인하였다.

• 비파괴검사학회지
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• 제24권2호
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• pp.164-170
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• 2004

중수로 내부구조물 중 칼란드리아관(CT)와 액체주입노즐관(LIN)은 서로 수평으로 90도 교차되게 배열되어 있으며 원자로 내의 열, 방사선, 하중에 의해 creep 현상이 발생되어 처짐이 일어난다. 칼란드리아관은 액체주입노즐관과 동일 재료이나 운전 온도와 방사선 조사량으로 인해 액체주입노즐관에 비해 상당히 열악한 조건에 노출되어 있으므로 처짐이 심각할 것으로 예상된다. 만약 두 관의 접촉이 발생되면 원전 안전성에 영향을 미칠 것이므로 인접관에 대한 접촉여부 점검은 중수로 안전현안 중 하나이다. 이러한 접촉여부를 확인하기 위하여 핵연료채널 내부로 탐촉자를 삽입하여 인접관과의 교차점에서 간격을 직접측정하기 위한 방법으로 원거리장 와전류검사 (RFECT) 기술을 적용하였다. 핵연료채널 인접관인 액체주입노즐관 신호 취득시 발생 가능한 잡음 신호(두께변화, Lift-off, 수축)에 대해 체적적분법에 의한 모델링으로 조사하였고, 신호와 잡음과의 분리 가능한 조건을 확인하였다. 원거리장 와전류검사 적정 조건은 민감도와 투과력 그리고 잡음신호 등을 동시에 고려하여 주파수 1kHz와 코일간격 200m로서 결정하였다. 원거리장 와전류검사 실험 결과 칼란드리아관과 액체주입노즐관 사이의 간격 변화에 대한 신호 특성을 전압평면을 이용하여 상관관계를 도출하였다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Journal of Advanced Marine Engineering and Technology
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• 제39권5호
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• pp.563-569
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• 2015

A chemical reaction occurring in CSTR (Continuous Stirred Tank Reactor) is significantly affected by the concentration, temperature, pressure, and reacting time of materials, and thus it has strong nonlinear and time-varying characteristics. Also, when an existing linear PID controller with fixed gain is used, the performance could deteriorate or could be unstable if the system parameters change due to the change in the operating point of CSTR. In this study, a technique for the design of a fuzzy PD plus I controller was proposed for the temperature control of a CSTR process. In the fuzzy PD plus I controller, a linear integral controller was added to a fuzzy PD controller in parallel, and the steady-state performance could be improved based on this. For the fuzzy membership function, a Gaussian type was used; for the fuzzy inference, the Max-Min method of Mamdani was used; and for the defuzzification, the center of gravity method was used. In addition, the saturation state of the actuator was also considered during controller design. The validity of the proposed method was examined by comparing the set-point tracking performance and the robustness to the parameter change with those of an adaptive controller and a nonlinear proportional-integral-differential controller.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.