• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.596-607
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• 2015

In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems.

• 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.21 no.1
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• pp.48-55
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• 1989

In this study, simple functional forms which could predict decay heat are referred to and modified in order to analyse more easily long-term behavior of decay heat generated from domestic PWR and CANDU spent fuel. To reduce the difference between the predicted data by functional forms and ORIGEN 2 results and to predict the decay heat under the important parameter(s), sensitivity analysis is performed. By introducing the identified hey parameter, turnup, into the functional forms, the decay heat of spent fuels within a limited rangs of cooling time(3~500 years) becomes predictable for various turnup rates. The predicted decay heat of spent fuels with representative turnup rates such as 33, 37 and 40 GWD/MTU by the functional forms is in so good agreement with ORIGEN 2 results within $\pm$10% difference over the cooling time from 1 to 10$^{5}$ years that the functional forms presented here may be used for engineering purposes such as the thermal design and assessment of the facilities associated with spent fuel management.

• Journal of Radiation Protection and Research
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• v.47 no.2
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• pp.99-106
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• 2022

Background: The radionuclide inventory calculation codes such as ORIGEN and FISPACT collapse neutron reaction libraries with energy spectra and generate an effective one-group cross-section. Since the nuclear cross-section data, energy group (g) structure, and other input details used by the two codes are different, there may be differences in each code's activation inventory calculation results. In this study, the calculation results of neutron-induced activation inventory using ORIGEN and FISPACT were compared and analyzed regarding radioactive waste classification and worker exposure during nuclear decommissioning. Materials and Methods: Two neutron spectra were used to obtain the comparison results: Watt fission spectrum and thermalized energy spectrum. The effective one-group cross-sections were generated for each type of energy group structure provided in ORIGEN and FISPACT. Then, the effective one-group cross-sections were analyzed by focusing on ⁵⁹Ni, ⁶³Ni, ⁹⁴Nb, ⁶⁰Co, ¹⁵²Eu, and ¹⁵⁴Eu, which are the main radionuclides of stainless steel, carbon steel, zircalloy, and concrete for decommissioning nuclear power plant (NPP). Results and Discussion: As a result of the analysis, ¹⁵⁴Eu and ⁵⁹Ni may be overestimated or underestimated depending on the code selection by up to 30%, because the cross-section library used for each code is different. When ORIGEN-44g, -49g, and -238g structures are selected, the differences of the calculation results of effective one-group cross-section according to group structure selection were less than 1% for the six nuclides applied in this study, and when FISPACT-69g, -172g, and -315g were applied, the difference was less than 1%, too. Conclusion: ORIGEN and FISPACT codes can be applied to activation calculations with their own built-in energy group structures for decommissioning NPP. Since the differences in calculation results may occur depending on the selection of codes and energy group structures, it is appropriate to properly select the energy group structure according to the accuracy required in the calculation and the characteristics of the problem.

• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.93-94
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• 2011

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.509-514
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• 1998

ORIGEN-S 전산코드로 계산된 가압경수로(PWR)사용후핵연료 내에 존재하는 방사성핵종비 $^{134}$ Cs/$^{137}$Cs 및 $^{154}$ Eu/$^{137}$Cs 를 감마선 분광실험으로 측정한 값과 비교하여 핵연료의 연소도를 결정하였다. 고리 1호기 및 2호기 사용후핵연료봉에 대한 감마선 분광실험을 한국원자력연구소 조사재시험시설(IMEF)과 조사후시험시설(PIEF)의 시험기기 및 장치를 이용하여 수행하고 이 결과로부터 $^{134}$ Cs/$^{137}$Cs 와 $^{154}$ Eu/$^{137}$Cs 의 핵종비를 측정하였다. 이와 별도로 사용후핵연료의 연소도, 냉각시간, 초기농축도등에 따른 $^{134}$ Cs/$^{137}$Cs 와 $^{154}$ Eu/$^{137}$Cs의 핵종비를 ORIGEN-S 코드로 계산을 하였으며, 이 핵종비와 연소도 사이의 관계를 회귀분석하여 2차 다항식 함수로 유도하였다 이관계식과 감마선 분광실험으로 측정한 $^{134}$ Cs/$^{137}$Cs와 $^{154}$ Eu/$^{137}$Cs 의 핵종비를 이용하여 각각의 연소도를 결정할 수 있었다.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.46-55
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• 1989

Neutron shielding for a spent fuel container was analized using the Monte Carlo code MCNP coupled with discrete ordinates codes, XSDRN and ONEDANT. The ORIGEN-S code was used to determine the fixed neutron source, and the spectral source information for MCNP were obtained from a 10 group XSDRN calculation and a 27 group ONEDANT calculation. When the depleted uranium shield was used, the results from ONEDANT and XSDRN calculations agreed with the MCNP results within 10% and 20%, respectively. Depleted uranium appears more effective than lead or steel as a neutron shield.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.63-72
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• 1989

An analysis for the gamma radiation fields in the research reactor MAPLE-X10 facility has been peformed under the assumption of partial loss of reactor and service pool water to assess the safety from the view point of design. Four photon source terms considered in the analysis were calculated using the ORIGEN-S code. Gamma dose rate calculations over the reactor and service pools during the water-loss accident conditions were performed using QAD-CG code. MCNP code (Monte Carlo Neuron and Photon Transport code), also, was used to assess the scattered radiation fields away from the pools, which is appropriate for calculating the scattered photon dose rates outside of the solid angle subtended by the source and pool walls.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.347-353
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• 2010

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.