• 대한기계학회논문집A
• /
• 제37권12호
• /
• pp.1573-1579
• /
• 2013

핵연료의 연소성능을 시험하기 위해서는 시험 루프에 설치된 조사리그 내에 냉각수가 순환되도록 설계되어야 한다. 이때, 조사리그 내 냉각수는 $300^{\circ}C$, 15.5 MPa 의 고온 고압으로 순환시키기 때문에 냉각수의 밀봉은 핵연료 조사리그를 제작할 때 가장 중요한 공정 중 하나이다. 특히 15 개의 계장선이 조사리그의 압력경계부위를 통과하게 되는데, 이의 밀봉을 위해 일반적으로 브레이징 공정이 적용된다. 본 연구에서는 조사리그 브레이징용 진공챔버 및 고주파 유도가열기를 포함하는 유도 브레이징 시스템을 개발하고, 다양한 실험을 통해 산화막이 발생하지 않는 공정변수를 검토하였으며, 브레이징 제품의 인장시험, 단면검사, 밀봉성능검사 등을 통해 브레이징 공정의 건전성과 밀봉성능을 검증하였다.

• 에너지공학
• /
• 제12권4호
• /
• pp.281-288
• /
• 2003

The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2002년도 하계학술대회 논문집 A
• /
• pp.265-267
• /
• 2002

This paper deals with an I/O simulator design for performance evaluation of reactor protection systems in nuclear power plants. The I/O simulator provides input signals for the reactor protection system, and acquires output signals from the initiation circuits. The simulator is based on VMEbus system, and all VMEbus boards are developed within the country.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.579-585
• /
• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Nuclear Engineering and Technology
• /
• 제52권11호
• /
• pp.2511-2521
• /
• 2020

The Fukushima accident is characterized by the fact that three reactors at the same site experienced reactor vessel failure and the accident resulted in significant radiological release to the environment, which was about 1/10 of the Chernobyl releases. The safe removal of fuel debris in the reactor vessel and Primary Containment Vessel (PCV) and treatment of huge amount of contaminated water are the major issues for the decommissioning in coming decades. Discussions on the new researches efforts being carried out in the area of investigation of the end state of fuel debris and Boling Water reactor (BWR) specific core melt progression, development of technologies for the mitigation of radiological releases to comply with the strengthened safety requirement set after the Fukushima accident are discussed.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• pp.651-656
• /
• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Nuclear Engineering and Technology
• /
• 제48권1호
• /
• pp.268-273
• /
• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• 한국물환경학회지
• /
• 제20권3호
• /
• pp.236-240
• /
• 2004

Autohydrogenotrophic denitrification of high nitrate concentration contaminated wastewater in a batch-scale biofilm reactor has been investigated. High nitrate concentration decreased as pH increased from 7.01 to 9.45. The high nitrate concentrations continuously decrease from $150mg.l^{-1}$ to $0mg.l^{-1}$. Nitrite concentrations increase at about two-thirds way through the denitrification process and thereafter it decreases with time. Autohydrogenotrophic denitrification of high nitrate concentration is passible to use drinking water as well as wastewater, and to deal with wastewater treatment by hetrotrophic denitrification.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
• /
• pp.291-295
• /
• 1996

In CANDU reactor, each pressure tubes contain twelve fuel bundles and provide the inlet and outlet for the primary coolant. If a leak develops in the pressure tube, it is detected by Annulus Gas System which contains circulating dry $CO_2$ gas. Since the leaks caused by the flaws are resulted in pressure tube break, establishment of flaw assessment method is very significant in view of the fracture mechanics. In this paper, various criteria for assessing the flaws are presented to prevent the tube rupture and ensure the integrity of reactor operating.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.484-488
• /
• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.