• Title/Summary/Keyword: KRR-2

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Analysis of Decommissioning Work Process for Thermal Column of KRR-2 (연구로 2 호기 Thermal Column 해체작업 공정분석)

  • 김성균;정운수;정기정
    • Proceedings of the Korea Society for Energy Engineering kosee Conference
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    • 2002.05a
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    • pp.125-128
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    • 2002

  • 국내 최초의 연구용 원자로 1,2 호기(KRR-1,2)는 각각 1962년과 1972년에 가동을 시작하여 국내 원자력 기술 자립에 많은 도움을 주었다. 그러나 하나로의 정상 가동으로 인해 연구용 원자로 1,2 호기의 효용가치가 상실되었을 뿐만 아니라 시설의 노후화로 인해 1995년 두기 모두 가동이 중지되었고 1996년 과학기술부에 의해 연구로 1,2 호기를 폐로 하기로 결정하였다.(중략)

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Derivation of site-specific derived concentration guideline levels at Korea Research Reactor-1&2 sites

  • Kim, Geun-Ho;Do, Tae Gwan;Kwon, Jae;Ryu, Gangwoo;Kim, Kwang Pyo
    • Nuclear Engineering and Technology
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    • v.54 no.2
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    • pp.493-500
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    • 2022

  • The objective of this study was to derive derived concentration guideline levels (DCGLs) reflecting the site-specific characteristics of KRR-1&2. A total of 7 nuclides (H-3, C-14, Co-60, Sr-90, Cs-137, Eu-152, and Eu-154) were selected for DCGLs derivation. Radiation dose at the sites was evaluated with RESRAD-ONSITE program. The dose contribution due to direct external exposure was the highest during the entire evaluation period. Ingestion had the second effect. The DCGLs of Co-60 was derived to be 0.051 Bq/g, and DCGLs of Cs-137 was 0.193 Bq/g. The DCGLs of H-3 showed the highest value of 129 Bq/g. The ratio of DCGLs derived by applying site-specific values and default values ranged from 0.27 to 19.6. For six nuclides excluding H-3, KRR-1&2 sites and the overseas NPP sites showed similar DCGLs. H-3 showed large differences in DCGLs from this study and overseas NPPs. The large difference resulted from input parameter values applied to the sites. In conclusion, it is critical to apply site-specific parameter values reflecting the site characteristics to derive DCGLs for decommissioned site clearance. The result of this study can be used as a reference for nuclide selection and DCGLs derivation reflecting the site characteristics when decommissioning nuclear facilities, including nuclear power plants in Korea.

  • https://doi.org/10.1016/j.net.2021.09.016 인용 PDF KSCI

Assessment of soil density and distribution coefficient of Cs-137 for deriving DCGLs in korea research reactor unit 1 and 2

  • Geun-Ho Kim;Ilgook Kim;Kwang Pyo Kim
    • Nuclear Engineering and Technology
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    • v.56 no.7
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    • pp.2452-2457
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    • 2024

  • To obtain site-specific values of the Derived Concentration Guideline Levels (DCGLs) for decommissioning of KRR-1&2, the soil density and distribution coefficient values for Cs-137, a major contaminant radionuclide, were determined. The soil density was evaluated according to the test method established by the Korean Agency for Technology and Standards of the Ministry of Trade, Industry, and Energy (KATS). The distribution coefficient was evaluated using a batch test. The validity of using the evaluated soil density and distribution coefficient as site-specific values was assessed through radiation dose assessment reflecting these values. Average soil density value obtained was 1.738 g/cm3, which was within the typical range of normal soil density, 1.0-1.8 g/cm3. The average distribution coefficient value was 7,754 mL/g. Applying the maximum, average, and minimum values of the evaluated soil density and distribution coefficient showed similar radiation dose results, thus suggesting that it is reasonable to use the average values of each parameter as site-specific values. Findings of this study can help determine DCGLs that reflect the characteristics of the research reactor site.

  • https://doi.org/10.1016/j.net.2024.02.002 인용 PDF

Internal Event Level 1 Probabilistic Safety Assessment for Korea Research Reactor (국내 연구용원자로 전출력 내부사건 1단계 확률론적안전성평가)

  • Lee, Yoon-Hwan;Jang, Seung-Cheol
    • Journal of the Korean Society of Safety
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    • v.36 no.3
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    • pp.66-73
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    • 2021

  • This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.

  • https://doi.org/10.14346/JKOSOS.2021.36.3.66 인용 PDF KSCI