• Nuclear Engineering and Technology
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• 제54권3호
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• pp.826-833
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• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.80-99
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• 2023

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10^th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• 대한토목학회논문집
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• 제5권3호
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• pp.21-29
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• 1985

활성(活性)슬러지가 팽화(膨化)를 하였거나 환기조내(環氣槽內)의 MLSS 농도(濃度)가 높은 경우, 종래(從來)의 중력침전법(重力沈澱法)으로는 슬러지의 분리(分離)가 어렵다. 이 문제(問題)의 해결책으로 가압부상법(加壓浮上法)을 이용(利用)하는 방법(方法) 연구(硏究)하였다. 가압부상법(加壓浮上法)의 효과(效果)에 영향(影響) 주는 인자중(因子中) 중요(重要)하다고 생각되는 환기조내(環氣槽內)의 MLSS 온도(溫度), 슬러지의 성상( 性狀), 가압수량비(加壓水量比), 압력(壓力)의 변화(變化)에 따른 가압부상법(加壓浮上法)의 효율(效率)에 관하여 회분식(回分式) 실험(實險)과 연속식(連續式) 실허(實驗)을 실시하여 다음과 같은 결과(結果)를 얻었다. 활성(活性)슬러지가 혼합액(混合液)의 분리(分離)가 종래(從來)의 중력침전법(重力沈澱法)으로는 어려운 경우에도 가압부상법(加壓浮上法)은 매우 좋은 효과(效果)를 나타낸다. 가압부상법(加壓浮上法)에는 한계(限界) 가압수량비(加壓水量比)가 존재(存在)하며 이 한계치(限界値)는 압력(壓力)에 따라 변화(變化)한다. 압력(壓力)은 단지 가압수량비(加壓水量比)에만 영향(影響)을 미치고 그 외(外)의 가압부상(加壓浮上) 효율(效率)에는 거의 영향(影響)이 없다. 연속식(連續式) 실험(實驗)이 회분식(回分式) 실험(實驗)보다 다소 효율(效率)이 떨어진다.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.104-106
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• 2005

In this work, a PWR reactor core dynamics is identified online by a recursive least squares method. Based on this identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to design an automatic controller for thermal power control in PWRs. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired one, and the variation of the control rod positions. Also, the objectives are subject to maximum and minimum control rod positions and maximum control rod speed. Therefore, the genetic algorithm that is appropriate to accomplish multiple objectives is used to optimize the model predictive controller. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), is used to verify the proposed controller for a nuclear reactor. From results of numerical simulation to check the performance of the proposed controller at the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are design requirements, it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1871-1885
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• 2019

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic P_N scattering, and the multi-group cross-section generation model was performed on the k_eff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a k_eff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 1997년도 추계학술대회 논문집
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• pp.617-620
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• 1997

In this paper, the thermal stress analysis of spent nuclear fuel disposal canister in a deep repository at 500m underground is done for the underground pressure variation. Since the nuclear fuel disposal usually emits much heat and radiation, its careful treatment is required. And so a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences some mechanical external loads such as hydrostatic pressure of underground water, swelling pressure of bentonite buffer, and the thermal load due to the heat generation of spent nuclear fuel in the basket etc.. Hence, the canister should be designed to designed to withstand these loads. In this paper, the thermal stress analysis is done using the finite element analysis code, NISA.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1603-1618
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• 2024

Emergency operation plans are indispensable elements for effective process safety management especially when unanticipated events occur under extreme situations. In the paper, a synthesis framework is proposed for the integration success path planning and performance analysis. Within the synthesis framework, success path planning is implemented through flow-directed signal tracing, renaming and reconstruction from a complete collection of Minimal Path Sets (MPSs) that are obtained using graph traversal search on GO-FLOW model diagram. The performance of success paths is then evaluated and prioritized according to the task complexity and probability calculation of MPSs for optimum action plans identification. Finally, an Auxiliary Feed Water System of Pressurized Water Reactor (PWR-AFWS) is taken as an example system to demonstrate the flow-directed MPSs search method for success path planning and performance analysis. It is concluded that the synthesis framework is capable of providing procedural guidance for emergency response and safety management with optimal success path planning under extreme situations.

• Journal of the Korean Wood Science and Technology
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• 제25권1호
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• pp.1-7
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• 1997

Kraft oak lignin and ricestraw lignin from acetosolve pulping were dissolved in 50/50 mixture of tetralin/m-cresol solvent. The dissolved lignin was reacted in the pressurized autoclave which was operating at $350{\sim}500^{\circ}C$ of reaction temperature and 10~20 atms of reaction pressure respectively_Hydrogen pressure of 60~80kg/$cm^2$ was exercising into the pressurized autoclave reactor to create thermochemical hydrogenolysis reaction. It was identified by GLC, GC-MS and HPLC that the alkyl-aryl-${\beta}$-O-4 ether bond of lignin was cleaved and degraded into various smaller molecules of aromatic compound such as phenols and cresols under the reaction conditions around $300^{\circ}C$ and 10 atms of reaction temoerature and pressure. Hydrogenolysis reaction of lignin compound which was done above $500^{\circ}C$ of reaction temperature and 20 atms of reaction pressure showed that the amount of aromatic compound such as phenols and cresols degraded from reactant lignin was decreasing with newly present and increasing water out of product mixtures. It was supposed that new aliphatic compound of high molecular weight hydrocarbon is composed due to higher reaction temperature and pressure of hydrogenolysis reaction such as $500^{\circ}C$ and 20 atms, even though it was almost impossible, to identify what kind of degraded products it was by HPLC.