• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• 원자력산업
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• 제17권3호통권169호
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• pp.4-9
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• 1997

70만kW급 가압 중수로형 원전(CANDU-PHWR)으로 건설되고 있는 월성 원자력 2호기가 오는 6월의 상업 운전을 목전에 두고 있다. 월성 2호기가 준공되면, 우리 나라의 원전 설비 용량은 1천만kW를 돌파하게 된다. 한국전력공사의 종합사업 관리하에 건설 추진되어 온 월성 2호기는 충분한 운전 경험을 통해 그 안전성과 신뢰성이 입증된 CANDU-600과 동일한 노형으로, 월성 1호기 건설 이후의 설비 개선과 최신 기술 기준 및 강화된 인허가 요건을 적용하여 발전소 안전성과 신뢰성을 제고하였으며, 선행 호기의 건설 경험을 최대한 활용하여 공정 계획을 수립$\cdot$추진하여 왔다. 그간의 건설 경위 등을 특징별로 살펴 본다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.502-507
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• 1998

Two different types the CANDU fuel bundles hue been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void paction distributions in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From calculated mixture enthalpy distribution at the exit of fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful assessing thermal behavior of the fuel bundle that could be used in CANDU reactors.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1892-1900
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• 2023

A protective oxide layer forms on the material surfaces of a Nuclear Power Plant during operation due to high temperature. These oxides can host radionuclides, the activated corrosion products of fission products, resulting in decommissioning workers' exposure. These deposited oxides are iron oxides such as Fe₃O₄, Fe₂O₃ and mixed ferrites such as nickel ferrites, chromium ferrites, and cobalt ferrites. Developing a new chemical decontamination technology for domestic CANDU-type reactors is challenging due to variations in oxide compositions from different structural materials in a Pressurized Water Reactor (PWR) system. The Korea Atomic Energy Research Institute (KAERI) has already developed a chemical decontamination process for PWRs called 'HyBRID' (Hydrazine-Based Reductive metal Ion Decontamination) that does not use organic acids or organic chelating agents at all. As the first step to developing a new chemical decontamination technology for the Pressurized Heavy Water Reactor (PHWR) system, we investigated magnetite dissolution behaviors in various HyBRID inorganic acidic solutions to assess their applicability to the PHWR reactor system, which forms a thicker oxide film.

• 한국가시화정보학회지
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• 제12권1호
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• pp.18-24
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• 2014

The Korea Atomic Energy Research Institute (KAERI) has a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. In the present work a preliminary experiment using a 1/8 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a velocity field is measured with a Particle Image Velocimetry (PIV) technique under an isothermal state. The flow patterns from the inlet nozzles to the top region of the tank are investigated using PIV for a 1/8 scale moderator tank.

• Nuclear Engineering and Technology
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• 제9권1호
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• pp.15-31
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• 1977

중수형 원자력발전소의 가동중에 연료를 재장전하는 특성을 고려하여 새로운 핵연료 batch와 주기의 개념을 서정하고, 연속적인 에너지 계산방법으로 개발하여 핵주기비 계산관계식을 유도하였으며, 이러한 관계식들로서 중수형 원자로에 사용될 수 있는 전자계산기 코드 HWRCOST를 개발하였다. 이 코드로서 현재 우리나라에 건설중인 CANDU-PHWR의 전수명에 걸친 핵연료 주기비를 계산하였고 아울러 우라늄 원광비, 성형 가공비, 사용핵연료 보관처리비 및 발전소 가동율의 변화에 대한 핵연료 주기비의 감응도를 분석하였다.

• Journal of Radiation Protection and Research
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• 제22권2호
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• pp.119-125
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• 1997

가압중수로는 감속재와 냉각재로 중수를 채택함으로써 높은 중성자 경제성을 달성하는 대신 중수소의 중성자 포획반응 때문에, 경수로에 비해, 다량의 삼중수소가 발생한다. 한편 원자로심에서, 삼중수소의 ${\beta}$-붕괴결과 발생된 $^3He$는, 열중성자를 포획하여 다시 삼중수소로 변환된다. 중수로에서 삼중수소의 생성에 대한 기존의 계산모형은, $^3He$가 상대적으로 낮은 용해도를 가지므로, 그 기여도를 무시해왔다. 그러나 $^3He$의 중성자 포획단면적은 중수소의 그것에 비해 $1.6{\times}10^7$ 배가 된다. 즉 $^3He$가 중수내에 0.03 ppm만 녹아있다 하더라도 $^3He$에 의해 생성되는 삼중수소의 양은 전체 중수에 의한 삼중수소의 양에 필적하게 된다. 본 연구에서는 월성1호기를 대상으로, 중수로에서 $^3He$가 삼중수소의 생성에 미치는 영향을 평가하였으며 결과를 실측치와 비교하였다. 연구의 결과, 감속재에서는 $^3He$의 용해도가 낮고 $^4He$ Cover gas 때문에 $^3He$의 기여도는 무시할 수 있음이 밝혀졌다. 반면 냉각재의 경우 $^3He$ 삼중수소의 생성에 지대한 영향을 미치는 것으로 나타났다. 또한 본 연구의 계산방법은 원전 운전초기의 냉각재내 삼중수소 생성량은 과대평가 하는 것으로 나타났으나 운전기간이 증가함에 따라 실측치와 잘 일치하는 것으로 나타났다.

• 한국가시화정보학회지
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• 제10권3호
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• pp.30-36
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• 2012

We are planning to construct a scaled-down moderator facility to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to investigate the anticipated problems of the flow visualization and measurement in the planning scaled-down moderator facility. We shortly describe CFD analysis result for the 1/40 scaled-down test model and the flow measurement techniques used for this test facility under isothermal flow conditions. The Particle Image Velocimetry (PIV) method is used to visualize and measure the velocity field of water in a transparent Plexiglas tank. Planar Laser Induced Fluorescence (PLIF) technique is used to evaluate the feasibility of temperature field measurement in the range of $20-40^{\circ}C$ of water temperature using an one-color method.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1448-1459
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• 2023

A key element of the safety analysis is Loss of Coolant Analysis (LOCA) which must be performed using system thermal-hydraulic codes. These codes are extensively validated against separate effect and integral experiments. RELAP/SCDAPSIM is one such code that may be used to predict LBLOCA response in a CANDU reactor. The RD-14M experiment selected for the Best Estimate Plus Uncertainty study is a 44 mm (22.7%) inlet header break test with no Emergency Coolant Injection. This work has two objectives first is to simulate pipe break with RELAP and compare these results to those available from experiment and from comparable TRACE calculations. The second objective is to quantify uncertainty in the fuel element sheath (FES) temperature arising from model coefficient as well as input parameter uncertainties using Integrated Uncertainty Analysis package. RELAP calculated results are found to be in good agreement with those of TRACE and with those of experiments. The base case maximum FES temperature is 335.5 ℃ while that of 95% confidence 95^th percentile is 407.41 ℃ for the first order Wilk's formula. The experimental measurements fall within the predicted band and the trends and sensitivities are similar to those reported for the TRACE code.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2003년도 춘계 학술발표회 논문집
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• pp.445-455
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• 2003

Relatively high carbon-14 emissions, which occurred at PHWR Plant during 1998 and 1999, made the site staff to implement several operational improvements: 1) the frequency and volume of the moderator cover gas purging were reduced through increased $O_2$ additions to the cover gas, 2) the 'old' resin columns were not used during re-start of the reactor from outage, 3) efforts were made to minimize air ingress, 4) the maximum service time of moderator ion-exchange columns were restricted to about 80 days. Through the improvements, the carbon-14 emission from each PHWR reactor returned to the normal levels during the remainder of 1999 and during 2000. We carried out a special surveillance at W-1 and W-3 from September 2001 to August 2002 to properly evaluate ways to optimize the use of moderator ion exchange resins from a C-14 perspective. The surveillance showed that only data that provided an operational marker for deciding when to remove the IX-resin column is an observed increase in the C-14 stack emissions themselves. Also, it is shown that any increase over the rate of 0.4 Ci $month^{-1}$ for two consecutive weeks may be the indication for an ion-exchange resin column change, especially if the IX-resin column has been in service for more than 80 days.