• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2699-2708
• /
• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 1998년도 하계학술대회 논문집 B
• /
• pp.813-815
• /
• 1998

Engineered Safety Features-Component Control Systems(ESF-CCS) are those I&C systems that control safety equipment used to maintain the integrity of reactor coolant pressure boundary. This paper illustrates distinctive features and improved design concepts of Korea Next Generation Reactor(KNGR) based on the experience obtained through prototyping of ESF-CCS.

• 한국산학기술학회논문지
• /
• 제14권5호
• /
• pp.2069-2074
• /
• 2013

내연기관 차량에 전자기식 클러치 워터펌프의 적용은 연비 향상 및 배기가스 저감을 꾀할 수 있다. 이러한 클러치 워터펌프는 엔진 냉각시스템의 유량 단속에 의하여 최적 운전 조건을 가능케 한다. 본 연구에서는 클러치 워터펌프를 이용한 냉각시스템을 제어함으로써 디젤 차량의 연비 및 배기가스 특성을 살펴보았다. 전자기식 클러치 워터펌프에 의한 저온 시동시 냉각수 흐름을 차단하여 아이들 조건에서 예열 시간을 기존 워터펌프 대비 49% 정도 단축시켰고, 주행 중에는 냉각수가 최적 고온상태를 유지하도록 제어하였다. 그리하여 NEDC 모드에서 연소 효율이 개선되어 최대 5% 정도의 연비 향상 효과를 나타내었다. 또한 NOx를 제외한 HC, CO 및 $CO_2$ 배기가스의 농도가 전반적으로 감소하였다.

• 한국자동차공학회논문집
• /
• 제19권5호
• /
• pp.58-66
• /
• 2011

A 2-loop waste heat recovery system with Rankine steam cycles for the improvement of fuel efficiency of gasoline vehicles has been investigated. A high temperature loop(HT loop) is a system to recover the waste heat from the exhaust gas, a low temperature loop(LT loop) is for heat recovery from the engine coolant cold relatively. This paper has dealt with a layout of a LT loop system, the review of the working fluids, and the design of the cycle. The design point and the target heat recovery of the LT boiler, a core part of a LT loop, has been presented and analytically investigated. Considering the characteristics of the cycle, the basic concept of the LT boiler has been determined as a shell-and tube type counterflow heat exchanger, the performance characteristics for various design parameters were investigated.

• Nuclear Engineering and Technology
• /
• 제45권5호
• /
• pp.573-580
• /
• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• 비파괴검사학회지
• /
• 제9권1호
• /
• pp.106-110
• /
• 1989

Bolting degradation problems in primary coolant pressure boundary applications have become a major concern in the nuclear industry. In the bolts concerned, the failure mechanism was either corrosion wastage(loss of bolt diameter) or stress-corrosion cracking.(3) Here the manual ultrasonic testing of RPV(Reactor Pressure Vessel) and RCP(Reactor Coolant Pump) stud has been performed. But it is difficult to detect indications because examiner can not exactly control the rotation angle and can not distinguish the indication from signals of bolt. In many cases, the critical sizes of damage depth are very small(1-2 mm order). At critical size, the crack tends to propagatecompletly through the bolt under stress, Resulting in total fracture.(3) Automatic stud scanner for studs(bolts) was developed because the precise measurement of bolt diameter is required in this circumstance. By use of this scanner, the rotation angle of probe was exactly controlled and the exposure time of radiations was reduced.

• 에너지공학
• /
• 제8권1호
• /
• pp.110-118
• /
• 1999

In order to protect the nuclear reactor coolant system from corrosion, lithium is injected into the coolant from the chemical injection tank. The present study investigates the chemical injection characteristics of the injection tank using a low Reynolds number turbulence model. Laminar flow analysis showed very little diffusion of the jet and gave incorrect flow and concentration fields. A disk located near the inlet of the injection tank was effective in mixing the chemical additives in the top portion of the tank, and significant reduction in injection time was obtained.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2000년도 추계학술대회 논문집 학회본부 C
• /
• pp.476-478
• /
• 2000

Objective is to design a high performance purification filter system of reactor coolant and seal injection system at nuclear power station. The purification filter systems play an important role in the stability of the nuclear and volume control system which consist the primary network systems of the nuclear power station. But the users of the purification filter systems frequently suffer from high maintenance cost which comes from lack of understanding of the system technology and domestic suppliers. It is time to establish a high performance domestic filter system manufacturing technology and optimum design for wide use in industrial applications.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• pp.695-700
• /
• 1997

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of .each strategy are also investigated.

• Nuclear Engineering and Technology
• /
• 제52권6호
• /
• pp.1110-1119
• /
• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.