• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.169-179
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• 1984

FIPREL 전산코드를 사용하여 원자로 냉각수 내의 핵분열 생성물에 의한 방사능을 분석함으로써 PWR의 운전시에 발생하는 핵연료 피복관 파손을 평가할 수 있는 효과적인 절차를 모색하였다. 이 코드를 이용하여 핵연료의 농축도, 연소도, 가동온도 및 갭유출계수의 크기로 정량화되는 실제적 파손 크기등의 물리적 파라미터에 대해서 핵분열 생성물의 방사능이 나타내는 민감도에 대한 방대한 계산을 실시하였으며 그 결과는 PROFIP방법에 의한 것과 대체적으로 일치한다. 노출 우라늄이 존재하는 경우에는 옥소보다도 화학적으로 더 안정된 핵종간의 방사능비에 근거하여 반복계산을 실시함으로써 파손된 핵연료 봉에서 유출된 방사능만을 분리해 낸다. 개발된 전산코드로 파손 핵연료봉의 선형출력 밀도, 갯수, 실제적 파손 크기 및 노출우라늄의 질량등을 계산할 수 있다. 고리 1호기의 4주기에 걸친 운전 경험을 이 모텔에 의해 분석한 결과에 의하면 본 모델은 원자력발전소 정상운전시 핵연료봉의 상태를 감시·평가하는데 아주 적합한 것으로 판명되었다.

• Journal of Radiation Protection and Research
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• 제14권2호
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• pp.30-36
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• 1989

고리 1호기의 사용후 핵연료에 함유된 핵종중 핵임계도에 미치는 영향이 큰 22개 핵종에 대한 핵단면적 자료를 XLACS-II에 의해서 ENDF/B-IV로 부터 취해서 FISSLIB(51군 핵단면적 자료)를 만들었다. 그리고 AMPX 조직에 의해서 DLC-43/CSRL 자료로 부터 생산된 51군 핵단면적 자료와 함께 사용할 수 있도록 하였다. 한편 BNL-325와 FISSLIB을 비교하여 후자의 유용성을 확인하였다. 그리고 FISSLIB에 수록된 핵분열 생성물 핵종에 대한 핵군단면적을 기본 입력자료로 하여 무한 배열된 고리 1호기 사용후 핵연료봉에 대한 핵임계도 계산을 수행하였다. 그 결과 냉각기간에 따라 핵분열 생성물 핵종을 고려할 때의 핵임계도는 그것을 무시할 때보다 9-14%의 감소효과를 나타냈다.

• 대한화학회지
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• 제45권4호
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• pp.304-311
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• 2001

경수로 사용후 핵 연료에 미량 함유되어 있는 핵분열생성물을 유도 결합 플라스마 원자방출분광법(ICP-AES)으로 분석하기 위하여 우라늄으로부터 학분열생성물을 추출 크로마토그래피로 분리, 회수하는 방법을 검토하였다. 우라늄 분리 분야에서 잘 알려져 있는 tri-n-butyl phosphate(TBP)를 추출제로 사용하여 몇 가지 Amberlite XAD 다공성 수지들에 대한 침윤능을 비교한 후 TPB침윤양이 가장 큰 Amberlite XAD-16을 지지체로 선택하였다. 사용후핵연료 용해용액과 화학조성이 유사한 모의 사용후핵연료 용해용액을 사용하여 TBP 침윤수지에 대한 핵분열생성물 원소들의 흡착거동을 조사하고, 분리에 미치는 여러 변수들을 최적화 하였다. Pd 및 Ru을 제외한 대부분의 핵분열생성물 원소들을 정밀도 3.1% 이하의 범위에서 정량적으로 회수할 수 있었다.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1300-1309
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• 2023

Simulated debris was synthesized using UO₂, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO₂, whereas a (U, Zr)O₂ solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U₃O₈ and (Fe, Cr)UO₄ phases formed at 1473 K, whereas a (U, Zr)O_2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.767-772
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• 2012

In support of optimizing electrorefining technology for treating spent nuclear fuel, lithium drawdown has been investigated for separating actinides from molten salt electrolyte. Drawdown reaction selectivity is a major issue that requires investigation, since the goal is to remove actinides while leaving the fission products and other components in the salt. A series of lithium drawdown tests with surrogate fission product chlorides was run to obtain selectivity data with non-radioactive salts, develop a predictive model, and draw conclusions about the viability of using this process with actinide-loaded salt. Results of tests with CsCl, $LaCl_3$, $CeCl_3$, and $NdCl_3$ are reported here. Equilibrium was typically achieved in less than 10 hours of contact between lithium metal and molten salt under well-stirred conditions. Maintaining low oxygen and water impurity concentrations (<10 ppm) in the atmosphere was observed to be critical to minimize side reactions and maintain stable salt compositions. An equilibrium model has been formulated and fit to the experimental data. Good fits to the data were achieved. Based on analysis and results obtained to date, it is concluded that clean separation between minor actinides and lanthanides will be difficult to achieve using lithium drawdown.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4084-4094
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• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.180-185
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• 1987

월성로형 핵연료봉의 정상상태 노내거동 분석용 전산코드인 KAFEPA를 개발하였다. KAFEPA 전산코드는 같은 목적하에서 AECL에 의해 개발되었던 ELESIM 전산코드에 대응하지만, KAPEPA 전산코드는 ELESIM에 비해 보다 이론적이고 정확하게 예측하는 계산모형들, 즉, 핵분열기체방출 모형, 노내고밀화모형 및 중성자속 감소 계산모형들을 내포하고 있다. KAFEPA 전산코드는 핵분열생성물 기체 방출에 대한 22개 노내 실험자료에 그 예측치를 비교함으로써 검증되었다. KAFEPA 전산코드에 의한 예측치는 상기 실험자료와 잘 일치하였다.