• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• 대한산업공학회지
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• 제35권2호
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• pp.178-184
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• 2009

The Nuclear Power Plant(NPP) industry in Korea has been making efforts to reduce the human errors which have largely contributed to about 150 nuclear reactor trips since 2001. Recently, the Crew Resource Management(CRM) training has risen as an alternative countermeasure against the nuclear reactor trips caused by human errors. The effectiveness of CRM training in NPP industry, however, has not been proven to be significant yet. In this study a return on investment(ROI) model is developed to measure the effectiveness of CRM training for the operators in Korean NPP. The model consists of mathematical expressions including multiple variables affecting the CRM training impacts and nuclear reactor trips. Implication of the model is discussed further in detail.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.536-547
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• 1994

본 논문에서는 최적제어기법을 이용한 원자로 출력 제어시스템을 다루었다. 시스템 변수들을 상태변수로 표시하면 관측치 뿐만 아니라 시스템 내부의 모든 상태변수를 동시에 다룰 수 있으므로 설계의 자유도가 증가될 수 있다. 따라서 본 논문에서는 원자로의 동특성식과 열수력학적 에너지 평형식을 사용하여 원자로를 모델링한 후 이를 상태변수로 나타내었다. 다음으로는 LQR 및 LQG 시스템을 설계하여 제어봉 및 출력의 거동을 동시에 만족시킬 수 있는 설계조건을 결정하였다. 또한 서보 시스템의 설계를 위해 보통의 휘드백 시스템과 차수를 증가시킨 레귤레이팅 시스템을 만들어 비교하였으며 그 결과 증가차수 레귤레이팅 시스템이 보통의 휘드백 시스템에 비해 우수한 제어 특성이 있음을 알 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.83-83
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• 2001

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2007년도 심포지엄 논문집 정보 및 제어부문
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• pp.253-254
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• 2007

This paper deals with the development of power control system(PCS) for nuclear power plants. The PCS provides the control motive power to operate the CEDMs(Control Element Drive Mechanism) for reactivity control inside the reactor vessel. The CEDM is raise and lower the CEAs( Control Element Assemblies) inside the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1479-1482
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• 2021

Research reactors are typically well-suited for outreach activities at different levels. However, unplanned seeking to increase the utilization of a research reactor may result in weakening the nuclear security of this facility. Research reactor staff might be in shortage of a functional nuclear security culture; specifically, there might be a conviction that the necessities of research can be given the priority over consistence with security procedural requirements. Research reactors are usually parts of bigger institutes or research labs of different activities. Moreover, the employments of research reactors are usually with the purpose that easy entry to the reactor premises is fundamental. So, they could be co-situated in places with different sorts of activities, mostly under similar security arrangements. The co-area of research reactor offices among different kinds of research labs introduces explicit security issues, the effects of which should be viewed as when building up a nuclear security framework. Notwithstanding potential security vulnerabilities presented in the design, research reactors frequently have devices kept promptly accessible to encourage research and education. The accessibility of these sorts of hardware could be used by an authorized person to commit an unapproved activity or cause harm. This paper aims to present insights to compromise contradicting conditions in new research reactors in which both enhancing utilization and ensuring security are satisfied.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.900-906
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• 2024

The Passively-Cooled Molten Salt Fast Reactor (PMFR) is one of the advanced design concepts of the Molten Salt Fast Reactor (MSFR) which utilizes a natural circulation for the primary loop and aims to attain a long-life operation without any means of fuel reprocessing. For an extended operation period, it is necessary to have enough fissile material, i.e., high excess reactivity, at the onset of operation. Since the PMFR is based on a fast neutron spectrum, direct implementation of a burnable absorber concept for the control of excess reactivity would be ineffective. Therefore, a localized moderator concept that encircles the active core has been envisioned for the PMFR which enables the effective utilization of a burnable absorber to achieve low reactivity swing and long-life operation. The modified PMFR design that incorporates a moderator and burnable absorber is presented, where depletion calculation is performed to estimate the reactor lifetime and reactivity swing to assess the feasibility of the proposed design. All the presented neutronic analysis has been conducted based on the Monte Carlo Serpent2 code with ENDF/B-VII.1 library.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3301-3312
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• 2023

This study incorporates a high-fidelity transient analysis solver based on multigroup CMFD in the MOC code STREAM. Transport modeling with heterogeneous geometries of the reactor core increases computational cost in terms of memory and time, whereas the multigroup CMFD reduces the computational cost. The reactor condition does not change at every time step, which is a vital point for the utilization of CMFD. CMFD correction factors are updated from the transport solution whenever the reactor core condition changes, and the simulation continues until the end. The transport solution is adjusted once CMFD achieves the solution. The flux-weighted method is used for rod decusping to update the partially inserted control rod cell material, which maintains the solution's stability. A smaller time-step size is needed to obtain an accurate solution, which increases the computational cost. The adaptive step-size control algorithm is robust for controlling the time step size. This algorithm is based on local errors and has the potential capability to accept or reject the solution. Several numerical problems are selected to analyze the performance and numerical accuracy of parallel computing, rod decusping, and adaptive time step control. Lastly, a typical pressurized LWR was chosen to study the rod-ejection accident.

• 대한전자공학회논문지
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• 제13권5호
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• pp.17-23
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• 1976

고속열중성자로에서 정상 운전중인 원자로를 운전정지하였다가 재가동할 때 가장 문제가 되는 것은 핵분열 생성물인 Xe135의 독소작용이다. 이것은 Xe135가 원자로 출력에 영향을 주는 열중성자에 대한 흡수단면적이 크고 그의 반감기가 길기 때문이다. 그러므로 원자로의 일시적 운전정지가 요구될 때 이의 재가시에는 반듯이 이 독소를 능과할 수 있는 충분한 초과반응도를 가해 주던지, Xe135가 붕괴되어 그의 농도가 줄어든 이후에야 원자로의 재가동이 가능하게 된다. 위와 같은 문제는 사실상 원자로 운전시 안전성 뿐만 아니라 경제성에도 큰 영향을 주고 있다. 본 논문에서는 이 점을 고려하여 Pontoyagin의 최대원리를 이용하여 운전정지를 최적화시키므로서 언제든지 원자로를 전출력으로 재가동할 수 있도록 운전정지 방법을 개선하였다. 그러나 제어과정에서나 그 이후에도 X, 농도는 제어된 허용치를 넘지 않고 최소시간 이내에 모든 제어를 끝내도록 하였다. The buildup of fission product, i.e. Xe-135 poisoning, is a prime factor in restarting a nuclear reactor from the shutdown, which was under normal operation in the high flux thermal reactor, It is caused by the high absorption crosssection of Xe-135 to thermal neutrons and its long half life, from which the thermal power is affected. It is then possible to restart a nuclear reactor after the sufficient excess reactivity to override this poisoning must be inserted, or its concentration is decreased sufficiently when its temporary shutdown is required. As ratter of fact, these have an important influence not only on reactor safety but also on economic aspect in operation. Considering these points in this study, the shutdown process was cptimized using the Pontryagin's maximum principle so that the shutdown mirth[d was improved as to restart the reactor to its fulpower at any time, but the xenon concentration did not excess the constrained allowable value during and after shutdown, at the same time all the control actions were completed within minimum time from beginning of the shutdown.