• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1360-1367
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• 2016

This paper introduces a boron meter with improved accuracy compared with other commercially available boron meters. Its design includes a new fitting function and a multi-detector. In pressurized water reactors (PWRs) in Korea, many boron meters have been used to continuously monitor boron concentration in reactor coolant. However, it is difficult to use the boron meters in practice because the measurement uncertainty is high. For this reason, there has been a strong demand for improvement in their accuracy. In this work, a boron meter evaluation model was developed, and two approaches were considered to improve the boron meter accuracy: the first approach uses a new fitting function and the second approach uses a multi-detector. With the new fitting function, the boron concentration error was decreased from 3.30 ppm to 0.73 ppm. With the multi-detector, the count signals were contaminated with noise such as field measurement data, and analyses were repeated 1,000 times to obtain average and standard deviations of the boron concentration errors. Finally, using the new fitting formulation and multi-detector together, the average error was decreased from 5.95 ppm to 1.83 ppm and its standard deviation was decreased from 0.64 ppm to 0.26 ppm. This result represents a great improvement of the boron meter accuracy.

• 비파괴검사학회지
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• 제24권4호
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• pp.307-314
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• 2004

배관의 축방향 균열 검사를 위하여 원주 방향으로 진행하는 유도초음파 모드를 적용하였다. 배관의 곡률을 변수로 원주 유도초음파의 분산선도를 계산하였으며 이를 배관 검사에 적용하기 위하여 중수로 피더관의 곡관부 축방향 균열을 탐지에 적용하였다. 상대적으로 낮은 주파수에서는 Lamb 파 특성을 따르나 주파수가 증가함에 따라 평판의 경우, 즉 곡률이 무한대인 경우 첫 번째 $A_0$ 모드와 두 번째 $S_0$ 모드가 합쳐져서 Rayleigh 모드 형태로 변화한다. 한편 곡률을 가진 배관의 경우는 주파수가 증가하더라도 첫 번째 모드와 두 번째 모드가 합쳐지지 않았다. 이러한 해석을 기초로 하여 배관의 일종인 중수로 피더관 곡관부 축방향 결함을 탐지하기 위하여 사각 탐촉자를 사용하여 Rocking 원주 유도초음파 기법을 적용하였다. 원주 방향으로 유도파를 진행시키면서 인공 결함으로부터의 수집된 신호를 분석하여 진동 모드를 확인하였으며 두께 대비 10% 깊이의 notch도 검출할 수 있음을 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제37권1호
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
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• 제13권1호
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• pp.1-11
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• 1981

가압경수로심의 3차원적 simulation 코드인 KINS를 개발하여 고리1호기 제 1주기에 대한 benchmark 계산을 수행하였다. KINS는 FLARE에서 사용하고 있는 모델을 기초로 하여 가압경수로심 해석에 보다 유용하게 쓸 수 있도록 발전시킨 것이다. 제 1주기초에서는 hot zero power 상태에서의 임계붕소농도, 핵연료집합체별 출력분포, 노심평균축방향 출력분포 등을 계산하여 실측 자료와 비교하였다. 아울러 연소도 1000MWD/MTU 단위로 연소계산을수행하여 여기서 산출된 임계 붕소농도와 핵 연료집합체별 출력 분포를 실측자료와 비교하였다. 계산결과는 실측자료와 매우 훌륭하게 일치하고 있으므로 KINS가 가압경수로의 노심관리에 아주 경제적이며 유효한 도구가 될것임을 보여주는 것이라고 생각된다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2104-2125
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• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1386-1395
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• 2020

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.645-660
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• 1995

표준원전을 대상으로하여 저수위 운전시의 잔열제거제통상실사고를 RELAP5/MOD3 및 RELAP5/MOD3.1 전산프로그램을 이용하여 분석하였다. 증기발생기가 이용가능할 때 원자로냉각재계통에 배기 경로가 없는 경우와 배기경로가 있는 경우에 대하여 분석을 수행하였다. 배기경로가 없는 경우에 대해 RELAP5 /MOD3 전산프로그램과 RELAP5 /MOD3.1 전산프로그램으로 비교 분석을 수행하였다. 분석 결과 두 전산프로그램의 계산결과는 정성적인 면 뿐 아니라 정량적 인면도 비교적 잘 일치하였다. 그러나 계산결과로부터 RELAP5 /MOD3의 경우에는 벽 열전달모델의 결함이 발견되어 배기경로가 있는 경우에 대해서는 RELAP5 /MOD3.1 전산프로그램을 이용하여 분석을 수행하였다. 분석결과 원자로정지후 하루가 지났을때 배기경로가 없는 경우에는 두개의 증기발생기로도 잔열이 충분히 제거되지 않아 원자로계통의 압력이 지속적으로 증가하여 사고개시 후4,000초 정도에 원자로계통의 임시밀봉재의 설계압력인 0.24MPa에 도달하였다. 가압기 안전밸브 용량의 세배정도 크기의 배기경로가 있는 경우에는 10,000 초가 지나도 원자로냉자재계통의 압력이 0.24 MPa에 도달하지 않았으며 노심노출이 초래되지 않았다. 분석결과의 상세한 검토를 통해서 저수위 운전시 잔열제거능력 상실사고가 발생하였을 경우 REL-AP5/MOD3.1을 이용한 사고해석 방법론의 타당성을 제안하였으며 또한 적절한 배기용량을 산정하기 위한 자료를 제공하였다.

• 한국산학기술학회논문지
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• 제10권11호
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• pp.3260-3268
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• 2009

냉각재상실사고이후 원전의 원자로건물집수조 여과기에서 화학적 영향을 고려한 수두손실을 종합적으로 평가하기 위한 시험장치를 개발하였다. 시험장치에서 원자로건물집수조와 시험장치에서 물 부피에 대한 여과기 면적의 비가 일치하도록 시험조건을 설정하고 시험을 수행하였다. TSP pH 조절제 조건에서 칼슘실리케이트는 시험 초기에 수두손실을 급격히 상승시켰기 때문에 원자로건물에서 모든 칼슘실리케이트를 제거하여야 함을 확인하였다. 비상노심냉각계통 살수지속시간의 차이에 따른 시험결과는 장기살수조건이 단기살수조건에 비해 12배 정도 높은 수두손실을 보였다. 살수조건 시험결과를 화학적 영향이 없는 수두손실과 비교하면 단기살수와 장기살수의 각 조건에서 5.6배 및 60.8배 수두손실이 증가하는 결과를 보였다. 화학적 영향은 재순환수에 노출된 물질의 양에 따라 초기의 일정기간 동안 알루미늄 및 아연도금 판의 부식에 의해 급격히 증가하고 이들이 부동피막을 형성한 이후에는 NUKONTM 및 콘크리트 등에서 침출된 화학종의 침전에 기인하여 증가율이 감소하는 경향을 보였다. 실험결과는 TSP에 의한 알루미늄의 부동피막 형성이 살수시간이 길어지고 알루미늄의 양이 많을 경우 효과적이지 않다는 것을 보였다.