• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.143-148
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• 1998

LCS와 ORIGEN2의 연결 프로그램 MONO를 개발하여 연소시간에 따른 가속기미임계로의 핵특성 변화를 분석할 있는 LCS-MONO-ORIGEN2 코드시스템을 구축하였다. 몇 가지 타입의 미임계로를 대상으로 LCS-MONORIGEN2 코드시스템의 성능시험을 수행하였다. 용융염 핵연료 및 집합체형 핵연료 미임계로에 대한 계산은 문제없이 수행되었다 또한 토륨/우라늄-233 핵연료 미임계로에 대한 연소시간에 따른 Keff 변화는 외국기관의 계산결과와 유사하게 나타났다.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.1-13
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• 1992

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup-dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORIGEN2-predicted burnup-dependent acti-nide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.415-415
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.393.2-393
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• 2002

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.69-75
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• 2004

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-1, To calculate neutron flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95% of the total activity in RPV came from the nuclides of $^{55}$ Fe, $^{60}$ Co, $^{59}$ Ni and $^{63}$ Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2% of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25${\times}$10$^{6}$ GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was peformed. The comparison results showed a good agreement.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.299-301
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.301-302
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• 2005

• Journal of Radiation Protection and Research
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• v.20 no.1
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• pp.53-64
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• 1995

The radiation fields following the large loss of coolant accident (LOCA) have been assessed for the vital areas in the service building of Wolsong 2, 3, and 4 nuclear power plants. The ORIGEN2 code was used in calculating the fission product inventories in the fuel. The source terms were based upon the activity released following the dual failure accident scenario, i.e., a LOCA followed by impaired emergency core cooling (ECC). Configurations of the reactor building, the service building, and the ECC system were constructed for the QAD-CG calculations. The dose rates and the time-integrated doses were calculated for the time period of upto 90 days after the accident. The results showed that the radiation fields in the vital access areas were found to be sufficiently low. Some areas however showed relatively high radiation fields that may require limited access.

• Proceedings of the KIEE Conference
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• 2000.11a
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• pp.262-264
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• 2000

원자로 내에서 연소 중인 핵연료나 저장 또는 재처리 중인 사용후핵연료의 성분으로서 시설의 공정설계, 안전성분석 및 차폐설계에 중요한 입력자료가 되는 핵분열생성물질, 방사화생성물 및 악티나이드의 핵종 농도와 이에 대응하는 방사능 강도의 기기 별 시간변 화율을 해석할 수 있는 코드 개발할 목적으로 MULTISAMS 정상 평형상태 모델을 구현하였다. MULTISAMS 코드의 반응공정 모델은 서로 연결되어 있으며 내부에 방사성물질의 혼합유체가 순환하는 세 종류의 반응기(원자로, 열교환기 및 화학반응기) 계통에서 자연적 또는 설계에 의해 일어나는 현상으로서; 반응기 간의 물질 흐름; 각 반응기 내에서 방사성 붕괴, 변환, 이동과 중성자 흡수 및 핵분열; 외부로부터 특정 핵종의 유입혹은 유출을 고려한 시간종속 핵종농도보존방정식 이론에 근거한다. 코드의 유용성 및 신뢰성을 검증하기 위해 현재 개념설계가 진행 중인 AMBIDEXTER원자력 에너지시스템을 대상으로 ORIGEN2 계산과 비교하였다. 두 코드 간의 입력조건과 배경이론차이점 때문에 절대적 비교가 불가능하므로 단순이론의 중간매개코드로서 SAMS를 이용한 2단계 비교방법을 따랐다. 결론은 MULTISAMS는 ORIGEN2 계산의 수렴치와 근사하게 일치하면서 ORIGEN2 가 다룰 수 없는 핵주기 연속후처리공정의 정상가동 시 핵종 평형농도를 기기 별로 계산할 수 있다는 장점을 확인하였다.