• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2004년도 추계학술발표대회 개요집
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• pp.67-69
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• 2004

The fuel assemblies as the nuclear fuel for the pressurized water reactor(PWR) are loaded in the reactor core throughout the residence time of three to five years. The spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the fuel assembly. The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral load acting on the fuel assembly so as to keep the fuel assembly straight. To meet the requirement, integrity on the spacer grid welding parts should be carefully checked. In this study, welding quality of the spacer grid assembly welded by several welding companies are examined and compared.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.420-424
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• 1994

가압경수로형 원자력발전소의 저출력 및 과도상태에서의 개선된 증기발생기 수위 제어 방식을 제시하였다. 수축 및 팽창 현상에 의한 수위의 요동을 줄이기 위해 기존의 비례·적분 제어기에 증기발생기 압력 및 급수온도를 고려한 앞먹임 보상부를 첨가하였다. 원전 훈련용 시뮬레이터를 이용하여 시뮬레이션을 수행한 결과 기존방식에 비해 적은 수위오차, 훨씬 빠른 진정시간을 얻을 수 있었다. 제시된 알고리즘은 구현이 용이하고 실제 적용도 가능하리라 판단된다.

• Nuclear Engineering and Technology
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• 제11권4호
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• pp.263-274
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• 1979

국내도입이 예상되는 900MWe급 가압경수로형 (PWR) 원자력 발전소와 캐나다형가압중수로형 (PHWR-CANDU) 원자력발전소에 대하여 throwaway 핵연료주기를 가상하여 두 노형의 상대적인 경제성을 비교 검토 하였다. 계산을 목적으로 발전단가를 발전소 투자비, 운전보수비, 운전자본비 및 핵연료비로 구분했으며 건설단가는 보완된 ORCOST 전산코드를 그리고 발전단가는 보완된 POWERCO-50 전산코드를 사용하여 구하였다. 계산에 요구되는 각종의 경제인자에 대하여는 단일의 수치값을 갖는 상수보다는 어떤 범위의 수치대를 이루는 통계적인 변수로 처리하였으며 ORCOST 및 POWERCO-50을 통한 무작위 추출법을 통하여 발전소 건설비 및 발전단가의 화율돗수 분포도를 얻었다. 계산결과 두노형간의 발전단가 분포도는 서로 겹치고 있으며 발전 단가의 기대치는 1986년도 미화로 PHWR의 발전단가가 PWR의 발전단가, 39.41mills/kwh보다 약 0.4mill/kwh만큼 적지만 PHWR의 건설기간이 PWR 보다 1년정도 더 걸리게되는 경우 차이가 없음을 알았다. 따라서 두 노형간의 경제성은 거의 우열을 가릴 수 없으며 한국에서 원자력발전소 노형을 선정할 때 기술전수, 국산화 등 경제외적 인자도 경제적 인자로 수량화하여 검토하는 것이 필요하다고 결론을 내렸다.

• 방사성폐기물학회지
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• 제12권3호
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• pp.199-209
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• 2014

가압경수로의 일차계통 제염을 위해 개발된 HYBRID 제염제의 재료부식 특성을 틈부식 시험방법을 사용하여 수행하였다. 기존 제염제의 부식특성과 비교하기 위하여 상용 제염제인 OA, CITROX 제염제의 부식특성도 함께 평가하였다. 시험재료는 가압경수로의 일차계통의 주 재료인 Alloy 600과 304 SS을 대상으로 시험하였다. 틈부식 시험은 가혹조건의 부식시험으로써 내식성이 강한 원전 구조재료의 건전성을 짧은 시간에 잘 확인할 수 있었다. 시험결과 OA와 CITROX 제염제에서는 crevice 시편 표면에 pitting과 IGA가 나타났으나 HYBRID 제염제에서는 국부부식이 전혀 발생되지 않았다. 무게감소 측정결과 HYBRID 제염조건에서는 $1.3{\times}10^{-3}{\mu}m/h$ 이하의 매우 낮은 부식속도를 나타내었다. 반면에, OA 제염제의 경우 Alloy 600은 $4.0{\times}10^{-2}{\mu}m/h$ 로 비교적 균일한 부식율을 나타내었으나, 304 SS의 경우 pH = 2.0 이하에서 급격한 가속부식을 나타내었다. HYBRID 제염제의 경우 일반부식에서뿐만 아니라 crevice 부식조건에서도 거의 부식이 일어나지 않아 PWR 계통제염 시 산화막 용해 후 제염제가 계통재료에 노출되어도 재료의 건전성이 입증되었다.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3171-3181
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• 2021

This research serves to advance the development of engineering computational fluid dynamics (CFD) computing efficiency for the analysis of pressurized water reactor (PWR) core using rod-type fuel assemblies with mixing vanes (one kind of typical PWR core). In this research, a CFD scheme based on the reconstruction of the initial fine flow field (RIF CFD scheme) is proposed and analyzed. The RIF scheme is based on the quantitative regulation of flow velocities in the rod-type PWR core and the principle that the CFD computing efficiency can be improved greatly by a perfect initialization. In this paper, it is discovered that the RIF scheme can significantly improve the computing efficiency of the CFD computation for the rod-type PWR core. Furthermore, the RIF scheme also can reduce the computing resources needed for effective data storage of the large fluid domain in a rod-type PWR core. Moreover, a flow-ranking RIF CFD scheme is also designed based on the ranking of the flow rate, which enhances the utilization of the flow field with a closed flow rate to reconstruct the fine flow field. The flow-ranking RIF CFD scheme also proved to be very effective in improving the CFD efficiency for the rod-type PWR core.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.