• Title, Summary, Keyword: CANDU

Search Result 520, Processing Time 0.05 seconds

Technical and Economic Evaluations of CANDU Advanced Fuel Bundle Designs (CANDU 개량 핵연료 설계 방안 분석)

  • Seok, Ho-Chun;Hwang, Wan;Park, Ju-Hwan;Kim, Bong-Gu;Sim, Ki-Sub;Jung, Chang-Jun;Heo, Y.H.;Jun, J.S.
    • Nuclear Engineering and Technology
    • /
    • v.22 no.4
    • /
    • pp.389-409
    • /
    • 1990
  • As a principal design of advanced CANDU fuel bundle, CANDU-KF39, CANDU-KF40 and CANDU-KF43 fuel bundles were proposed and evaluated with respect to the operating conditions of the CANDU-6 reactor of Wolsung Unit-1. From the results, the advanced fuel bundles show to be improved economical and technical benefits compared with the current 37-element bundle. Especially, it was appeared that the KF-39 fuel bundle has more benefits of the safety, technical and economical aspects of Wolsung Unit-1 rather than those of the KF-40 and KF-43 fuel bundles.

  • PDF

Evaluation of CANDU Pressure Tube Integrity (CANDU 압력관의 건전성 평가)

  • 지세환;김영진
    • Journal of the KSME
    • /
    • v.33 no.5
    • /
    • pp.449-455
    • /
    • 1993
  • 지금까지의 CANDU 사고이력과 관련된 문제점을 살펴보면 핵연료 채널의 부적절한 설계 및 설치 그리고 부적절한 압력관 가동조건 등에 많은 문제점이 있었다. 이러한 의미에서 CANDU의 안전성은 압력관의 건전성으로부터 확보된다 하여도 과언이 아니다. 그러나 CANDU에서 차지 하는 중요성에 비추어 압력관의 사용환경은 매우 열약하다. 따라서 가동중 압력관 건전성 위협 요인에 대한 정기적인 검사, 시험 및 평가는 CANDU 안전성확보의 첫걸음이 된다. 특히 건전 성평가에 필요한 주요자료가 압력관 인출시험결과로부터 확보됨을 고려할 때 압력관 인출시험을 국내에서 수행할 수 있는 능력을 확보하는 것 또한 우리에게 부과된 과제라 할 것이다.

  • PDF

Assessments of RELAP5/MOD3.2 and RELAP5/CANDU in a Reactor Inlet Header Break Experiment B9401 of RD-14M

  • Cho Yong Jin;Jeun Gyoo Dong
    • Nuclear Engineering and Technology
    • /
    • v.35 no.5
    • /
    • pp.426-441
    • /
    • 2003
  • A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

Transient Analysis of the CANDU-9 480/SEU Reactor (CANDU-9 480/ SEU 원자로의 과도변화해석)

  • J. C. Shin;Park, J. H.;K. N. Han;H. C. Suk
    • Nuclear Engineering and Technology
    • /
    • v.27 no.5
    • /
    • pp.687-700
    • /
    • 1995
  • The thermal-hydraulic transient analysis of the proposed CANDU-9 plant was peformed. Several major transients ore analyzed if they meet the heat transport system design requirements. The proposed heat transport system configuration and the preliminary sizes of system equipment are justified by analysis in terms of the fuel integrity and the high system pressure limit during transients. The compliance with AECB R-77 requirements for CANDU-9 reactor was estimated. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements and the fuel overheating is prevented. One pump start-up during the reactor start-up operation was analyzed to investigate the How reversal through the fuel channel, which is specific in the proposed CANDU-9 plant.

  • PDF

Radiation Shielding Calculation on Shield System of CANDU 6 Plant Using the Coupled DOT4.2 and QAD-CG Codes (DOT4.2-QAD-CG 접속법을 이용한 CANDU 6 발전소 차폐 계통에 대한 방사선 차폐 계산)

  • Kim, Kyo-Youn;Kim, Jong-Kyung
    • Nuclear Engineering and Technology
    • /
    • v.25 no.4
    • /
    • pp.561-569
    • /
    • 1993
  • DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6 $\mu$Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.

  • PDF

Radiation Shielding Analysis of CANDU Spent Fuel Transport Cask (CANDU 사용후핵연료 수송용기 방사선차폐 영향평가)

  • Choi, Jong-Rak;Yoon, Jung-Hyun;Kang, Hee-Young;Lee, Heung-Young;Chung, Sung-Whan
    • Journal of Radiation Protection and Research
    • /
    • v.18 no.2
    • /
    • pp.27-35
    • /
    • 1993
  • A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

  • PDF

A Dynamic Model of U-Tube Steam Generator for CANDU Simulation

  • Lim, Jae-Cheon;Seoungyon Cho
    • Proceedings of the Korean Nuclear Society Conference
    • /
    • /
    • pp.213-218
    • /
    • 1996
  • A simulation model for the transient behavior of CANDU U-tube steam generator(UTSG) has been developed for application to the simulation of operational transient behavior of CANDU nuclear power plant. For application to CANDU UTSG. tile design characteristics of CANDU UTSG such as Wolsong Units, feedwater inlet near the tube sheet. is approximated. For realistic prediction of thermal hydraulic behavior of and tube bundle region is divided into two separate control volumes, subcooled region and saturated region. and the variation of thermal hydraulic properties within a control volume is considered. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator and considered to be applicable to the simulation of overall plant.

  • PDF