• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.346-347
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• 2017

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.241-242
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 학술논문집
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• pp.174-174
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• 2004

It is necessary to perform the radionuclide inventory assessment for the disposal of low and intermediate radioactive waste containers. The $\gamma$ nuclide analyzer can be used for the assessment of containers. However, if the radioactivity in the containers is extremely low or high, radionuclide inventory of the containers can not be evaluated properly. Also, gamma scanning method is time consuming and has economical burden to the utilities.(omitted)

• 방사성폐기물학회지
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• 제12권2호
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• pp.171-178
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• 2014

원전 해체 시 원자력설비는 안전하게 해체되어야 한다. 고리 1호기나 월성 1호기와 같은 노후화된 원전의 경우 곧 원전 해체를 계획하고 있는 대상 원전이지만, 이 원전들의 가동 중단 후 해체 시 선원항 평가 기준, 제염 및 해체 기술 등의 독자적인 국내 기술 확보는 미흡한 실정이다. 본 연구의 목적은 원전 선원항 평가 기술 중 하나로 In-Situ 기법을 이용하여 대형 원전 기기를 직접 측정하여 측정대상체에 대한 선원항 평가방법을 개발하는 것이다. 원자로 헤드를 별도의 해체 없이 이동형 감마핵종분석기를 이용하여 직접 측정법으로 분석하고 간접 측정을 병행하여 측정 결과를 보완하였다. 그리고, 표면오염시료는 방사화학분석을 수행하였다. 분석 결과를 확장하여 원자로의 핵종 재고량을 계산하였다. 본 연구 결과를 토대로 각 핵종별 방사능량 변화에 따라 해체 시점을 결정할 수 있으며, 원전 해체 시 작업자의 피폭 저감에 도움이 될 것으로 기대한다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1907-1916
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• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1226-1233
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• 2017

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. The database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

• 방사성폐기물학회지
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• 제2권1호
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• pp.35-40
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• 2004

Final disposal of radioactive waste generated from Nuclear Power Plant (NPP) requires the detailed information about the characteristics and the quantities of radionuclides in waste package. Most of these radionuclides are difficult to measure and expensive to assay. Thus it is suggested to the indirect method by which the concentration of the Difficult-to-Measure (DTM) nuclide is estimated using the correlations of concentration - it is called the scaling factor - between Easy-to-Measure (Key) nuclides and DTM nuclides with the measured concentration of the Key nuclide. In general, the scaling factor is determined by the log mean average (LMA) method and the regression method. However, these methods are inadequate to apply to fission product nuclides and some activation product nuclides such as 14$^{C}$ and 90$^{Sr}$ . In this study, the artificial neural network (ANN) method is suggested to improve the conventional SF determination methods - the LMA method and the regression method. The root mean squared errors (RMSE) of the ANN models are compared with those of the conventional SF determination models for 14$^{C}$ and 90$^{Sr}$ in two parts divided by a training part and a validation part. The SF determination models are arranged in the order of RMSEs as the following order: ANN model

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• 방사성폐기물학회지
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• 제14권2호
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• pp.91-100
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• 2016

경수로 원전을 대상으로 원전 내 방사화 대상 물질인 스테인리스강, 탄소강 및 콘크리트의 불순물 정보 적용여부에 따른 방사화 핵종 재고량을 계산하였다. 본 연구에서 탄소강은 압력용기 물질에 사용되었고, 스테인리스강은 압력용기 내부 물질에 사용되었으며, 일반 콘크리트가 생체 차폐체에 사용되었다. 금속 물질에 대해서는 참고자료 1개의 불순물 함량 정보를 적용하였고, 콘크리트 물질에서는 참고자료 5개의 불순물 함량 정보를 적용하여 평가를 수행하였다. 방사화 핵종 재고량 전산해석 시 중성자속 계산에는 MCNP 전산코드를, 방사화 계산에는 FISPACT 전산코드를 각각 사용하였다. 계산 결과, 금속 물질에서 불순물을 포함한 경우가 그렇지 않은 경우보다 비방사능이 2배 이상 높았으며, 특히 콘크리트에서는 불순물을 포함한 경우가 그렇지 않은 경우보다 최대 30배 이상 비방사능이 높게 계산되었다. 방사화 핵종의 생성반응과 재고량을 분석한 결과, 금속 구조물에서는 불순물 중 Co원소와 중성자에 의해 생성되는 방사화 핵종인 Co-60이, 콘크리트에서는 불순물 중 Co, Eu 원소와 중성자에 의해 생성되는 방사화 핵종인Co-60, Eu-152, Eu-154 이 방사성폐기물 준위 결정에 큰 영향을 미치고 있음을 확인하였다. 본 연구의 결과는 원전 해체 계획 수립 시 방사화 핵종 재고량 평가 및 규제에 활용될 수 있을 뿐 아니라, 해체를 고려한 원전 또는 원자력시설의 설계 단계에서도 참고자료로 활용 될 것으로 판단된다.

• Journal of Radiation Protection and Research
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• 제47권2호
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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.