• 제목/요약/키워드: Reactor module

검색결과 145건 처리시간 0.038초

CSPACE for a simulation of core damage progression during severe accidents

  • Song, JinHo;Son, Dong-Gun;Bae, JunHo;Bae, Sung Won;Ha, KwangSoon;Chung, Bub-Dong;Choi, YuJung
    • Nuclear Engineering and Technology
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    • 제53권12호
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    • pp.3990-4002
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    • 2021
  • CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

Study on the influence of flow blockage in severe accident scenario of CAP1400 reactor

  • Pengcheng Gao;Bin Zhang ;Jishen Li ;Fan Miao ;Shaowei Tang ;Sheng Cao;Hao Yang ;Jianqiang Shan
    • Nuclear Engineering and Technology
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    • 제55권3호
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    • pp.999-1008
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    • 2023
  • Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

막결합형 고온 이상 혐기성 소화공정에서 음폐수 처리 특성 (Characteristics of Food Waste Leachate Treatment in Thermophilic two Stage Anaerobic Digestion Combined UF Membrane)

  • 김영오;전덕우
    • 유기물자원화
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    • 제20권3호
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    • pp.21-24
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    • 2012
  • 본 연구에서는 폐기물 자원화 공정에서 발생하는 음폐수를 UF 분리막을 결합한 혐기성 소화 파일럿 에서 처리하면서 혐기성 소화효율, 바이오가스 생산량과 운전인자를 도출하였다. 운전방식은 막내부에서 외부로 고액분리가 이루어지는 가압식이었으며, 투과 플럭스는 15∼20 LMH, 막간 차압은 $1{\sim}3 kgf/cm^2$였다. 분리막 운전은 직접순환방식으로 운전한 결과, 메탄조의 수질악화로 간접순환방식으로 운전하였다. 유입수의 TCOD 와 SCOD 는 각각 113 g/L, 62 g/L 이었고, 유출수의 TCOD 와 SCOD는 각각 25 g/L, 12 g/L 이었다. TCOD 및 SCOD 제거효율은 각각 77% 및 81%였다. 하지만 UF 공정을 추가했을 때는 제거효율이 93% 및 86%로 증가하였다.

디젤탈황 단위공정의 CFD 모델링을 포함한 연료전지 시스템 공정설계 및 최적화 (Process Simulation and Optimization of Fuel Cell System including CFD Modeling of Diesel Desulfurizer Unit Process)

  • 최창용;임도진
    • Korean Chemical Engineering Research
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    • 제56권3호
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    • pp.421-429
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    • 2018
  • 본 연구에서는 100 kW급 연료전지 시스템의 운영을 위한 공정 및 CFD 모델링을 진행하였다. 공정 모델링을 통해 연료전지 각 단위 공정에 유입되는 유량을 도출하였으며 수소로 전환되지 않는 디젤의 환류량을 도출하였다. 디젤의 환류를 고려한 새로운 유입 유량 조건을 이용해 CFD 해석을 진행한 결과, 환류 디젤이 없는 것으로 가정한 이전 연구결과에 비해 더 짧은 유입시간과 비슷한 시간의 처리시간을 가지는 이점이 있음을 확인하였다. 6기의 탈황 반응기를 이용해 100 kW급 연료전지를 가동시키는데 필요한 TSA 탈황 시스템 구성을 완료하였으며 전체 TSA 공정 운영을 위한 운용 방안을 도출하였다. 반응기 사이의 열 전달 해석을 통해 저온의 탈황공정과 고온의 재생공정 간의 열 간섭이 크지 않음을 확인하였다. 본 연구결과는 연료전지 시스템의 효율화에 기여할 것이며, 도출된 탈황모듈의 설계는 연료전지 시스템뿐만 아니라 청정 석유화학산업의 기초가 될 것으로 기대된다.

RADIOLOGICAL DOSE ASSESSMENT ACCORDING TO METHODOLOGIES FOR THE EVALUATION OF ACCIDENTAL SOURCE TERMS

  • Jeong, Hae Sun;Jeong, Hyo Joon;Kim, Eun Han;Han, Moon Hee;Hwang, Won Tae
    • Journal of Radiation Protection and Research
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    • 제39권4호
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    • pp.176-181
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    • 2014
  • The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.

Modification of RFSP to Accommodate a True Two-Group Treatment

  • Bae, Chang-Joon;Kim, Bong-Ghi;Suk, Soo-Dong;D. Jenkins;B. Rouben
    • 한국원자력학회:학술대회논문집
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    • 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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    • pp.185-190
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    • 1996
  • RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the *SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

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Establishment of DeCART/MIG stochastic sampling code system and Application to UAM and BEAVRS benchmarks

  • Ho Jin Park;Jin Young Cho
    • Nuclear Engineering and Technology
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    • 제55권4호
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    • pp.1563-1570
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    • 2023
  • In this study, a DeCART/MIG uncertainty quantification (UQ) analysis code system with a multicorrelated cross section stochastic sampling (S.S.) module was established and verified through the UAM (Uncertainty Analysis in Modeling) and the BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) benchmark calculations. For the S.S. calculations, a sample of 500 DeCART multigroup cross section sets for two major actinides, i.e., 235U and 238U, were generated by the MIG code and covariance data from the ENDF/B-VII.1 evaluated nuclear data library. In the three pin problems (i.e. TMI-1, PB2, and Koz-6) from the UAM benchmark, the uncertainties in kinf by the DeCART/MIG S.S. calculations agreed very well with the sensitivity and uncertainty (S/U) perturbation results by DeCART/MUSAD and the S/U direct subtraction (S/U-DS) results by the DeCART/MIG. From these results, it was concluded that the multi-group cross section sampling module of the MIG code works correctly and accurately. In the BEAVRS whole benchmark problems, the uncertainties in the control rod bank worth, isothermal temperature coefficient, power distribution, and critical boron concentration due to cross section uncertainties were calculated by the DeCART/MIG code system. Overall, the uncertainties in these design parameters were less than the general design review criteria of a typical pressurized water reactor start-up case. This newly-developed DeCART/MIG UQ analysis code system by the S.S. method can be widely utilized as uncertainty analysis and margin estimation tools for developing and designing new advanced nuclear reactors.

고분자 중공사막 모듈을 이용한 미세기포 발생과 이미지 분석기법을 이용한 기포 특성 파악 (Utilization of Image Analysis Technique for Characterization of Micro-Bubbles Generated by Polymeric Membrane Module)

  • 김준영;장인성
    • 대한환경공학회지
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    • 제33권6호
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    • pp.447-452
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    • 2011
  • 본 연구에서는 고분자 중공사막 모듈을 산기 도구로 사용하여 미세 기포 발생을 도모하였고 영상정보를 기반으로 하는 이미지 분석 기법을 적용하여 발생한 기포 특성을 파악하고자 하였다. 기포의 이미지 분석 결과, 고분자 중공사막 모듈을 통해 수중에 분사된 기포는 산기석을 통한 기포보다 약 30~64% 작은 크기로 발생되는 것을 관찰하였고, 기포의 70% 이상이 0.2~0.82 mm 범위에 분포된 반면, 산기석의 경우는 0.77~1.08 mm의 범위에 속한 기포가 80% 이상이었다. 산기석과 고분자 중공사막모듈을 각각 기포발생 장치로 사용한 부상조를 운영하였을 때 반응조에 잔존하는 플록의 크기는 산기석을 이용했을 때가 고분자 중공사막 모듈의 경우보다 더 큰 것으로 나타났다. 이는 고분자 중공사막 모듈에 의해 발생한 미세기포가 충돌 효율의 증가 때문에 크기가 작은 콜로이드 입자들까지 응집/부상할 수 있는 기회를 제공하였기 때문이다. 따라서 부상공정에 고분자 중공사막모듈을 산기장치로 활용하였을 경우 응집부상 제거효율이 증가할 수 있는 가능성을 보였다. 또한 본 연구에서 사용한 이미지 분석기법은 제공된 영상정보를 기반으로 기포의 기초 특성들과 관련된 정보습득을 위한 분석도구로 활용될 수 있을 것으로 판단된다.

Pilot Scale Plant의 황 충진 MBR을 이용한 고효율의 질소제거 공법 개발 (Development of High-rate Nitrogen Removal Process Using Submerged MBR Packed with Granular Sulfur of Pilot Scale Plant)

  • 문진영;황용우;조현정
    • 상하수도학회지
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    • 제25권3호
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    • pp.383-390
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    • 2011
  • In this study, a process combined biofiltration with sulfur-utilizing autotrophic denitrification and membrane separation was proposed to examine the efficiency of nitrogen removal. As an experimental device, hollow-fiber module was installed in the center of reactor to generate the flux forward sulfur layer in the cylinder packed with granular sulfur. In addition, a simple module was installed in activated sludge aeration tank which inside and outside of sulfur-using denitrification module was covered with microfilter and the module was considered as an alternative of clarifier. The experiment for developing new MBR process was carried out for three years totally. As the results of first two-year experiment, successful nitrogen removal performance was revealed with lab-scale test and pliot scale plant using artificial wastewater and actual plating wastewater. In this year, pilot scale test using actual domestic wastewater was performed to prove field applicability. As the results, high-rate nitrogen removal performance was confirmed with about 0.19 kg ${NO_3}^--N/m^3$ day of rate. Also significant fouling and pressure increase were not found during the experiment. And, the production ratio of sulfate and the consumption ratio of alkalinity showed a slightly higher value about 311 mg ${SO_4}^{2-}/L$ and 369 mg $CaCO_3$/L, respectively. In conclusion, the developed MBR process can be utilized as an alternative for retrofiting existing wastewater plants as well as new construction of advanced sewage wastewater treatment plants, with cost-effective merit.

동시 질산화-탈질(SND) 반응을 적용한 MBR 반응조에서 질소 및 인 제거 특성 (Nitrogen and Phosphorus Removal in Membrane Bio-Reactor (MBR) Using Simultaneous Nitrification and Denitrification (SND))

  • 전동걸;임현숙;안찬현;이봉규;전항배;박찬일
    • 대한환경공학회지
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    • 제35권10호
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    • pp.724-729
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    • 2013
  • 동시 질산화 탈질은 미세 용존 산소하에 한 반응조내에서 일어난다. 따라서, 본 연구에서는 인 방출을 위해 공기가 공급되는 MBR 전단에 혐기성 존을 만들어주었으며, 높은 DO 농도에서 탈질효율을 향상시켜 주기 위해서는 MBR 내에 배플을 설치하여 무산소 존이 이루어지게 하였다. 그리고 인 제거를 위한 테스트는 MBR 전단의 혐기성 반응조에 알럼 응집제를 투입하여 수행하였다. 질소 제거를 위한 SND의 최적 DO 농도 도출은 MBR 내 DO 농도를 2.0, 1.5, 1.0, 0.75 mg/L의 다양한 조건에서의 운전을 통해 수행하였다. 심지어 높은 알칼리성 하수라 알럼 응집제를 투입하였을 때 알칼리 용액 첨가 없이도 pH는 7.0~8.0로 유지되었다. TCODcr와 $NH_4^+$-N의 제거 효율은 모든 DO 농도에서 90% 이상이었다. DO 농도 2.0, 1.5, 1.0, 0.75 mg/L에서의 TN 제거효율은 각각 50, 51, 54, 66%이었다. DO 농도 0.75 mg/L 조건에서 알럼을 첨가한 결과 TN 제거효율은 54%로 감소하였다. 혐기성 반응조에 알럼을 투입한 결과 TP 제거효율은 29%에서 95%로 향상되었다. 그리고 알럼 투입 후 분리막 모듈의 화학적 세정 주기는 15~20일부터 40~50일으로 늘어났다.