• Nuclear Engineering and Technology
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• 제50권6호
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• pp.899-906
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• 2018

The irradiation performance of $U_3Si$ dispersion fuel in an Al matrix, $U_3Si-Al$, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized $U_3Si$ powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1195-1208
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• 2019

This paper introduces a new two-step procedure for PWR depletion analyses. This procedure adopts the albedo-corrected parameterized equivalence constants (APEC) method to correct the lattice-based raw cross sections (XSs) and discontinuity factors (DFs) by accounting for neutron leakage. The intrinsic limitations of the conventional two-step methods are discussed by analyzing a 2-dimensional SMR with the commercial DeCART2D/MASTER code system. For a full-scope development of the APEC correction, the MASTER nodal code was modified so that the group constants can be corrected in the middle of a microscopic core depletion. The basic APEC methodology is described and color-set problems are defined to determine the APEC functions for burnup-dependent XS and DF corrections. Then the new two-step method was applied to depletion analyses of the SMR without thermal feedback, and its validity was evaluated in terms of being able to predict accurately the reactor eigenvalue and nodal power profile. In addition, four variants of the original SMR core were also analyzed for a further evaluation of the APEC-assisted depletion. In this work, several combinations of the burnup-dependent and -independent XS and DF corrections were also considered. The results show that the APEC method could enhance the nodal equivalence significantly with inexpensive additional costs.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1355-1364
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• 2019

A recently published nuclear forensics methodology for source discrimination of separated weapons-grade plutonium utilizes intra-element isotope ratios and a maximum likelihood formulation to identify the most likely source reactor-type, fuel burnup and time since irradiation of unknown material. Sensitivity studies performed here on the effects of random measurement error and the uncertainty in intra-element isotope ratio values show that different intra-element isotope ratios have disproportionate contributions to the determination of the reactor parameters. The methodology is robust to individual errors in measured intra-element isotope ratio values and even more so for uniform systematic errors due to competing effects on the predictions from the selected intra-element isotope ratios suite. For a unique sample-model pair, simulation uncertainties of up to 28% are acceptable without impeding successful source-reactor discrimination. However, for a generic sample with multiple plausible sources within the reactor library, uncertainties of 7% or less may be required. The results confirm the critical role of accurate reactor core physics, fuel burnup simulations and experimental measurements in the proposed methodology where increased simulation uncertainty is found to significantly affect the capability to discriminate between the reactors in the library.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.369-376
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• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.35-40
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• 1995

Considering the higher discharge burnup, lower channel refuelling rate, lower linear element rating(LER), lower coolant void reactivity and axial power shape, CANFLEX-RU fuel bundle is optimized for CANDU-6 by grading the fissile composition in the ring-wise of the bundle and by applying fuel management scheme appropriately. The fissile composition of the fuel bundle is graded as the recovered uranium (0.9 w/o U-235) in the outer and intermediate elements, depleted Uranium (0.2 w/o U-235) in the center element, natural uranium (0.71 w/o U-235) in the inner elements. Enrichment is not required for these fuel. The fissile composition is optimized by lattice calculation and by time-averaged reactor simulation. CANFLEX-RU optimized for CANDU-6 resulted to be the 15% lower channel refuelling rate, acceptable axial power profile and power envelope, 70% higher discharge burnup, 15% lower LER and not increase coolant void reactivity compared with the 37-element natural uranium bundle for CANDU-6.

• 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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• 제55권6호
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

• 방사성폐기물학회지
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• 제9권2호
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• pp.73-80
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• 2011

한국방사성폐기물관리공단 주관 하에 개념 설계된, 연소도이득효과 적용 대용량 수송용기에 대해 방사선 차폐 안전성을 평가하였으며 여러 방사선원들이 수송용기 주변 선량률 분포에 미치는 영향을 분석하였다. 가능한 모든 방사선원(중성자선원, 감마선원, 방사화선원)들을 고려하였으며 보수적인 가상의 핵연료(너비: WH 17 RFA, 축방향: CE Type)를 선정, 실제 상황과 동일한 조건이 되도록 계산모델을 구축하였다. 모든 조건(정상 및 가상사고 조건)에서 표면선량률과 외부선량률이 법적기준치를 만족하고 있었으며 축방향 높이에 따라 각 선원들의 기여도가 변하고 있었지만 정상조건에서의 최대 표면선량률과 외부선량률은 방사화선원에 의한 영향이 가장 높은 것으로 확인되었다. 가상사고 조건에서는, 중성자선원의 선량률 기여도가 대략 90%에 달하고 있었으나 수송용기 끝단에서는 방사화선원에 의한 선량률이 급격하게 상승함에 따라 BUC 적용 수송용기의 방사선 차폐해석시 충분히 보수적으로 해석되도록 방사화선원을 정밀하게 분석하여 설정하여야 할 것으로 판단되었다.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2000년도 추계 학술발표회 논문집
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• pp.105-109
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• 2000

종래의 핵연료소결체가 혼합산화물 혹은 이산화우라늄중 한가지 핵연료만으로 구성한 것과 달리 내부를 저농축 이산화우라늄 핵연료로 채우고 그 외부를 링형태의 혼합산화물 핵연료로 둘러 싼 이중구조를 특징으로 한다. 이러한 형태의 핵연료소결체는 중심영역의 핵분열반응률 줄임으로써 핵분열 기체생성, 핵연료봉 중심온도와 평균온도를 낮추어 준다. 이는 핵분열 기체방출을 낮추어 혼합산화물 핵연료봉 성능을 향상시키고 방출 연소도를 증가시키는 효과가 있다.