본문 바로가기
HOME> 저널/프로시딩 > 저널/프로시딩 검색상세

저널/프로시딩 상세정보

권호별목차 / 소장처보기

H : 소장처정보

T : 목차정보

Journal of the Korean Nuclear Society = 원자력학회지 12건

  1. [국내논문]   Development of a Mechanistic Fission Gas Release Model for LWR $UO_2$ Fuel Under Steady-State Conditions  

    Koo, Yang-Hyun ; Sohn, Dong-Seong
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 229 - 246 , 1996 , 0372-7327 ,

    초록

    A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO $_2$ fuel. Under the assumption that UO $_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140 $0^{\circ}C$ , wide scatter band for fuel with centerline temperatures lower than 100 $0^{\circ}C$ , and underprediction for fuel with centerline temperatures higher than 140 $0^{\circ}C$ .

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  2. [국내논문]   A Study on the Implementation Effect of Accident Management Strategies on Safety  

    Jae, Moo Sung ; Kim, Dong Ha ; Jin, Young Ho
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 247 - 256 , 1996 , 0372-7327 ,

    초록

    A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO $_2$ fuel. Under the assumption that UO $_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140 $0^{\circ}C$ , wide scatter band for fuel with centerline temperatures lower than 100 $0^{\circ}C$ , and underprediction for fuel with centerline temperatures higher than 140 $0^{\circ}C$ .

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  3. [국내논문]   Feasibility of Long Term Feed and Bleed Operation For Total Loss of Feedwater Event  

    Kwon, Young-Min ; Song, Jin-Ho
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 257 - 264 , 1996 , 0372-7327 ,

    초록

    The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  4. [국내논문]   Loss of a Main Feedwater Pump Test Simulation Using KISPAC Computer Code  

    Jeong, Won-Sang ; Sohn, Suk-Whun ; Seo, Ho-Taek ; Seo, Jong-Tae
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 265 - 273 , 1996 , 0372-7327 ,

    초록

    Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  5. [국내논문]   Effect of Exchangeable Cation on Radionuclide Diffusion In Compacted Bentonite  

    Choi, Jong Won ; Park, Hyun Soo ; Oscarson, D. W.
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 274 - 279 , 1996 , 0372-7327 ,

    초록

    Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  6. [국내논문]   Effect of Bundle Junction Face and Misalignment on the Pressure Drops Across a Randomly Loaded and Aligned 12 Bundles in Candu Fuel Channel  

    Suk, H. C. ; Sim, K. S. ; Chung, C. H. ; Lee, Y. O.
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 280 - 289 , 1996 , 0372-7327 ,

    초록

    Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  7. [국내논문]   A Study on Effect of Capture Volume in a Cavity on Direct Containment Heating Phenomena  

    Chung, C.Y. ; Kim, M.H. ; Lee, H.Y. ; Kim, P.S.
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 290 - 298 , 1996 , 0372-7327 ,

    초록

    Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  8. [국내논문]   Onset of Slugging Criterion Based on Singular Point and Stability Analyses of Transient One-Dimensional Two-Phase Flow Equations of Two-Fluid Model  

    Sung, Chang-Kyung ; Chun, Moon-Hyun
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 299 - 310 , 1996 , 0372-7327 ,

    초록

    A two-step approach has been used to obtain a new criterion for the onset of slug formation : (1) In the first step, a more general expression than the existing models for the onset of slug flow criterion has been derived from the analysis of singular points and neutral stability conditions of the transient one-dimensional two-phase flow equations of two-fluid model. (2) In the second step, introducing simplifications and incorporating a parameter into the general expression obtained in the first step to satisfy a number of physical conditions a priori specified, a new simple criterion for the onset of slug flow has been derived. Comparisons of the present model with existing models and experimental data show that the present model agrees very closely with Taitel & Dukler's model and experimental data in horizontal pipes. In an inclined pipe ( $\theta$ =50<TEX> 50 $^{\circ}$ ), however, the difference between the predictions of the present model and those of existing models is appreciably large and the present model gives the best agreement with Ohnuki et al.'s data.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  9. [국내논문]   On the Reconstruction of Pinwise Flux Distribution Using Several Types of Boundary Conditions  

    Park, C. J. ; Kim, Y. H. ; Cho, N. Z.
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 311 - 319 , 1996 , 0372-7327 ,

    초록

    A two-step approach has been used to obtain a new criterion for the onset of slug formation : (1) In the first step, a more general expression than the existing models for the onset of slug flow criterion has been derived from the analysis of singular points and neutral stability conditions of the transient one-dimensional two-phase flow equations of two-fluid model. (2) In the second step, introducing simplifications and incorporating a parameter into the general expression obtained in the first step to satisfy a number of physical conditions a priori specified, a new simple criterion for the onset of slug flow has been derived. Comparisons of the present model with existing models and experimental data show that the present model agrees very closely with Taitel & Dukler's model and experimental data in horizontal pipes. In an inclined pipe ( $\theta$ =50<TEX> 50 $^{\circ}$ ), however, the difference between the predictions of the present model and those of existing models is appreciably large and the present model gives the best agreement with Ohnuki et al.'s data.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지
  10. [국내논문]   Extensions of Streaming Rays Method for Streaming Dominant Neutron Transport Problems  

    Hong, Ser-Gi ; Cho, Nam-Zin
    Journal of the Korean Nuclear Society = 원자력학회지 v.28 no.3 ,pp. 320 - 330 , 1996 , 0372-7327 ,

    초록

    The streaming rays(SR) method is improved and extended to multigroup, anisotropic scattering, and three-dimensional angular space(x-y-z(infinite))problems. This method is applied to the shielding problems in which the ray effect occurs seriously. For verification, the results of MORSE-CG code are used as reference solution and the results of TWODANT code are compared. The results show that solutions of the SR method are much better than those of the TWODANT code and are in good agreement with those of the MORSE-CG code. Also, to reduce computing time, two acceleration algorithms are implemented in the SR method : the standard coarse-mesh rebalance and a new angular two-grid acceleration.

    원문보기

    원문보기
    무료다운로드 유료다운로드

    회원님의 원문열람 권한에 따라 열람이 불가능 할 수 있으며 권한이 없는 경우 해당 사이트의 정책에 따라 회원가입 및 유료구매가 필요할 수 있습니다.이동하는 사이트에서의 모든 정보이용은 NDSL과 무관합니다.

    NDSL에서는 해당 원문을 복사서비스하고 있습니다. 아래의 원문복사신청 또는 장바구니담기를 통하여 원문복사서비스 이용이 가능합니다.

    이미지

    Fig. 1 이미지

논문관련 이미지