• Nuclear Engineering and Technology
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• 제52권1호
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• pp.14-18
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• 2020

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

• Journal of Radiation Protection and Research
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• 제18권2호
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• pp.47-59
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• 1993

본 연구에서는 사용후핵 연료를 안전하게 수송할 수 있는 수송용기의 여러 가지 설계 항목중에 수송용기 내부에 장전한 핵연료에 의한 핵임계반응을 방지하기 위한 핵임계해석을 수행하였다. 핵임계 해석에 사용한 HANSEN-ROACH-KENO-Va 전산시스템에 대한 검증계산을 수행하였고 수송용기의 핵임계측면에서의 안전성을 확보하기 위해 가능한 보수적인 가정을 하여 어떠한 경우에도 수송용기에 장전된 핵연료가 임계상태에 도달하지 않도록 수송용기 내부의 구조 및 적절한 핵임계 방지제를 선택하였고 정상수송 및 가상사고 조건 등에 대한 해석을 수행하였다. 그 결과 KSC-7 수송용기 의 설계조건을 만족하고 핵임계측면에서의 안전성을 보장할 수 있는 재료 및 구조에 대한 결론을 해석적으로 도출하였다.

• Journal of Radiation Protection and Research
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• 제30권2호
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• pp.61-67
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• 2005

노심설계 해석체계로 사용하는 HELIOS-MASTER코드를 이용하여 SAMRT연구로 사용 후 핵연료 저장조에 대한 임계도 해석체계를 개발하였다. 저장조의 기하학적 모형에 대한 거시 단면적을 HELIOS코드를 이용하여 생산하고, 저장조의 3차원 모델에 대한 임계도를 MASTER코드로 평가하였다. 또한 3차원 MCNP계산을 통하여 HELIOS-MASTER체계를 이용한 임계도 평가의 타당성을 검증하였다 HELIOS-MASTER코드 체계를 이용한 임계도 해석결과가 약간 보수적인 방향으로 허용오차 범위 내에서 정확도를 유지하였다. HELIOS-MASTER 코드 체계는 3차원 연소계산이 가능하기 때문에 차후에 연소이력을 고려한 사용후 연료 저장조에 대한 임계해석에 유용할 것이다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2010년도 학술논문요약집
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• pp.271-272
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• 2010

Sensitivity analysis by TSUNAMI clarifies the complex effects of key nuclides on the criticality probability quantitatively. As discussed above, the $K_{eff}$ of $UO_2$ fuel reaches the maximum value with 42w% concentration of intrusion water. The concentration of hydrogen affects the complexity of reaching criticality by its competition between the concentrations of $^{235}U$. Approximately if the weight percent of $H_2O$ in the mixture is less than 42%, the moderation effect of hydrogen surpasses its dilution effect on $^{235}U$. However, the importance of $^{235}U$ increases dramatically when the weight percent of water is bigger than 42%. In the sensitivity evaluation of $UO_2$ fuel employing TSUMAMI, there is a similar crosspoint of the sensitivity of $^{235}U$ and the sensitivity of $^1H$ where the criticality reaches summit. And the optimal water weight percent is determined to be 50%.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.249-253
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• 2001

Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of $^{24}Na$ in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of $^{24}Na$ is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 $[(Bq^{24}Na/g^{23}Na)/mGy]$.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2009년도 학술논문요약집
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• pp.321-322
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• 2009

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• Journal of Radiation Protection and Research
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• 제9권2호
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• pp.97-102
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• 1984

KSC-1 핵연료(核燃料) 수송용기(輸送容器)에 대(對)한 핵림계분석(核臨界分析)을 KENO-IV 몬테칼로 전산(電算)코드와 AMPX 전산(電算)코드계(系)로 부터 생산(生産)한 CSLIB 19 19-에너지군(群) 단면적(斷面積) 자료(資料)를 써서 수행(修行)하였다. 이때 미국(美國) B&W 사(社) CX-10 핵림계장치(核臨界裝置)를 대상으로 하여 KENO-IN 및 CSLIB 19단면적(斷面積) 시스템에 대한 검증계산(檢證計算)을 수행(遂行)한 후(後), 이 시스템의 타당성을 먼저 확인(確認)하였다. 핵림계분석(核臨界分析) 결과(結果), 1개(個)의 가압경수로(加壓輕水爐) 사용후(使用後) 핵연료집합체(核燃料集合體)를 운반할 수 있는 핵연료수송용기(核燃料輸送容器)는 정상적(正常的)인 수송조건(輸送條件)뿐만 아니라 가상적(假想的)인 수송사고조건하(輸送事故條件河)에서도 핵림계(核臨界)에 관(關)한 한(限) 안전(安全)한 것 같았다.