• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1896-1906
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• 2020

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark calculations were performed by DeCART stand-alone and DeCART/MATRA multi-physics coupled code system to verify their accuracy. The solutions of DeCART stand-alone calculations for the control rod bank worth, detector signal, isothermal temperature coefficient, and critical boron concentration agreed very well with the measurements. The root-mean-square errors of the boron letdown curves for two-cycles were less than about 20 ppm, while the individual and total control rod bank worth agreed well within 7.3% and 2.4%, respectively. For the BEAVRS benchmark calculations at the beginning of burnup, the difference between DeCART simplified thermal-hydraulic stand-alone and DeCART/MATRA coupled calculations were not significantly large. Therefore, it is concluded that both the DeCART stand-alone code and the DeCART/MATRA multi-physics coupled code system have the capabilities to generate high fidelity transport solutions at core follow calculations.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2130-2137
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• 2020

The standard method for calculating anti-neutrino emissions from a reactor involves knowing the fractional fission rates for the most important fissioning nuclides in the reactor. To calculate these rates requires detailed reactor physics calculations based upon the reactor design, fuel design, burnup dependent fuel composition, location of specific fuel assemblies in the core and detailed operational data from the reactor. This has only been published for a few reactors during specific time periods, whereas to be of practical use for anti-neutrino reactor monitoring it is necessary to be able to predict these on the publicly available information from any reactor, especially if using these data to subtract the anti-neutrino signal from other reactors to identify an undeclared reactor and monitor its operation. This paper proposes a method to estimate the fission fractions for a specific reactor based upon publicly available information and provides a database based upon a series of spent fuel inventory calculations using the FISPIN10 code and its associated data libraries.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1998.05a
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• pp.221-228
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• 1998

2차원 노심핵설계 코드 HELIOS를 이용하여 $^{7}$ LiF-BeF$_2$-ThF$_4$-$^{233}$ UF$_4$ 용융염 핵연료와 흑연(Graphite) 감속재로 구성된 AMBIDEXTER(Advanced Molten-salt Break-even Inherently-safe Dual-mission EXperimental and TEst Reactor) 원자로의 육각주형 로심격자에 대해 핵적 자활성 요건의 설계해석을 수행하였다. AMBIDEXTER 원자로는 액체 핵연료의 유동성을 이용한 온라인 핵연료 정화ㆍ처리ㆍ재생의 연속공정을 도입하여 노내의 잔류 핵분열 생성물질의 포화양을 최소로 유지시키고 중성자 경제성을 극대화하므로 높은 전환율을 얻는 설계이다. 핵연료 내에 잔류하는 핵분열생성물질의 포화농도에 대응하는 연소도를 등가연소도로 정의할 때, 열출력 250MW$_{th}$ AMBIDEXTER 원자로의 등가연소도 374MWD/TeH.E.의 평형 로심 모델에 대해 핵적 자활성을 지배하는 주요 핵설계 인자로서 용융염 핵연료의 $^{233}$ U Mole 분율, 흑연-대-용융염의 체적비, 노심격자 간격 및 출력 밀도의 변화에 따른 임계도 및 전환율을 평가하였다. 그 결과, $^{233}$ U Mole 분율과 혹연-대-용융염 체적비를 좌표축으로 하는 2차원상공간에서 핵적 자활성 요건 상태함수는 각 노심격자간격에 대해 완만한 선형 함수로 표현할 수 있음을 확인하였다.

• Korean Journal of Materials Research
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• v.14 no.7
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• pp.451-456
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• 2004

Thermal creep properties of the zirconium tube which was developed for high burnup application were evaluated. The creep test of cladding tubes after various final heat treatment was carried out by the internal pressurization method in the temperature range from $350^{\circ}C to 400^{\circ}C$ and from 100 to 150 MPa in the hoop stress. Creep tests were lasted up to 900days, which showed the steady-state secondary creep rate. The creep resistance of zirconium claddings was higher than that of Zircaloy-4. Factors that affect creep resistance, such as final annealing temperature, applied stress and alloying element were discussed. Tin as an alloying element was more effective than niobium due to solute hardening effect of tin. In case of advanced claddings, the optimization of final heat treatment temperature as well as alloying element causes a great influence on the improvement of creep resistance.

• Nuclear Engineering and Technology
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• v.12 no.4
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• pp.274-285
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• 1980

The TDCORE and the RELOAD-II codes are adopted in an attempt to establish a simplified computational system for the in-core fuel management decisions of a PWR. The TDCORE is being used to simulate the power and burnup behavior of the Gori unit 1 PWR during the fuel cycle 1 through the cycle 5. The validity of the TDCORE code is also presented by comparing the TDCORE prediction with the in-core measurements. The RELOAD-II code is used to determine the optimum fuel loading pattern which is one of the most important decisions of the Gori unit 1 reactor. The utility and applicability of two codes for the fuel management analyses are described.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.332-340
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• 1989

Destructive examination of 14$\times$14 PWR fuel burned for 3 cycles are carried out to investigate the in-reactor fuel performance. The results obtained are as follows; 1) Grain growth is not occured at the fuel center. 2) Fuel density is decreased as the turnup increase, the density is down to 94.4% TD at burnup of 36,000 MWD/MTU. 3) Average thickness of oxide layer on cladding is less than 10 $\mu$m in the lower and middle section, while it is rapidly increased above 20 $\mu$m in the upper section. 4) The rate of hydride production in the cladding is large in the upper section than lower section and is related to the production of oxide on the cladding

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.245-258
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• 1989

A 69-group cross section library consisting of more than 130 materials was generated for thermal reactor applications using the NJOY nuclear data processing system and the recent version of evaluated nuclear data files available from IAEA Nuclear Data Section. The multigroup library was validated through the analysis of various criticality experiments and depletion results of PWR. When used with the WIMS-KAERI code, the average $K_{eff}$ obtained for 47 uranium-oxide and 41 uranium metal fueled critical configurations is 0.9997 with a standard deviation of 0.69 percent. The calculated burnup dependent isotopic inventories of uranium and plutonium generally show good agreement with measured values obtained from depleted PWR pins.s.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.298-306
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• 2001

As fuel burnup proceeds, thermal gradients, differential swelling, and inter-assembly loading may induce assembly duct bowing. Since duct bowing affects the reactivity, such as long or short term power-reactivity-decrement variations, handling problem, caused by top end deflection of the bowed assembly duct, and the integrity of the assembly duct itself. Assembly duct bowing were first observed at EBR-ll in 1965, and then several designs of assembly ducts and core restraint system were used to accommodate this problem. In this study, NUBOW-2D KMOD was used to analyze the bowing behavior of the assembly duct under the KALIMER(Korea Advanced Liquid MEtal Reactor) core restraint system conditions. The mechanical behavior of assembly ducts related to several design parameters are evaluated. ACLP(Above Core Load Pad) positions, the gap distance between the ducts, and the gap distance between the duct and restraint ring were selected as the sensitivity parameter for the evaluation of duct deflection.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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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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• v.52 no.3
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• pp.538-543
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• 2020

Boric acid is used as a coolant for pressurized-water reactors, and the degree of burnup is controlled by the concentration of boric acid. Therefore, accurate measurement of the concentration of boric acid is an important factor in reactor safety. An improved system was proposed for the accurate determination of boron concentration. A new boron-concentration measurement technique, called multisensitive detection, was developed to improve the measurement accuracy of boron meters. In previous studies, laboratory-scale experiments were performed based on different sensitivity detectors, confirming a 65% better accuracy than conventional single-detector boron meters. Based on these experimental results, an experimental system simulating the coolant-circulation environment in the reactor was constructed; accuracy analysis of the boron meter with a multisensitivity detector was performed at the actual coolant pressure and temperature. In this study, the boron concentration conversion equation was derived from the calibration test, and the accuracy of the boron concentration conversion equation was examined through a repeatability test. Through the experiment, it was confirmed that the accuracy was up to 87.5% higher than the conventional single-detector boron meter.