• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.70-75
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• 1997

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14$\times$14 PWR fuel assembly and the results are given in this paper.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.58-63
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• 1998

AMBIDEXTER(Advanced Molten-Salt Break even Inherently-Safe Dual-Mission EXperiment & TEst Reactor)는 토륨-우라늄 연료주기의 핵적자활성 요건을 설계하는 방법으로써 핵분열중간 생성물인 $^{233}$ Pa의 시간격리, 노내 방사성물질 농도저감, 잉여반응도 및 증식률향상을 위해 핵분열 생성물질의 온라인 정화.처리.재생 개념을 채택하고 있다. 본 연구에서는 AMBIDEXTER 로심의 핵분열성물질의 연소와 온라인 정화.처리에 따른 핵연료내 원소분포 변화를 기술하기 위해 핵분열생성물질의 평형포화농도에 대응하는 등가연소도(Equivalent Burnup)를 정의하고 이를 노심의 핵적자활성 요건에 대해 최적화하는 핵연료 정화공정의 시간상수 특성을 시뮬레이션 하였다. 핵분열생성물질농도의 동특성은 ORIGEN2 코드에 내장된 연속재처리 모델을 이용하여 해석하였으며 실용화가 입증된 후보정화공정들을 고려하여 모든 핵종을 5종의 핵종군으로 분류하여 평가하였다. 시뮬레이션 결과 유효정화주기를 0.1 (노심장전량/일)로 연속재처리 할 때 노심내 포화등 가연소도는 약 650 (MWD/TeH.E.)로 대응되며 이때 동일한 핵연료량으로부터 생성된 노내 핵분 열생성물질 평형농도는 최대연소도 33000MWD/TeU의 PWR 평형노심 BOC시의 대비해 약 1/10 에 해당하는 양이 잔유하는 것으로 나타났다.

• 대한전기학회논문지:시스템및제어부문D
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• 제54권9호
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• pp.567-569
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• 2005

NARX(Nonlinear AutoRegressive with eXogenous input) neural network was used for prediction of nuclear reactor behavior which was influenced by control rods in short-term period and also by the concentration of xenon and boron in long-term period in load following operations. The developed model was designed to predict reactor power, xenon worth and axial offset with different burnup states when control rods and boron were adjusted in load following operations. Data of the Korea Next Generation PWR were collected by ONED94 code. The test results presented exhibit the capability of the NARX neural network model to capture the long term and short term dynamics of the reactor core and the developed model seems to be utilized as a handy tool for the use of a plant simulation.

• Nuclear Engineering and Technology
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• 제10권1호
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• pp.26-36
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• 1978

Hot cell의 설계를 위하여 기사용 연료에서의 방사능과 붕괴 에너지의 수치적 계산을 하였다. 고리 1호기와 같은 경수로에서 거의 최대 연소율인 33,000MWD/T(U)으로 태워진 연료봉 시험을 위하여 보관할 수 있는 최적의 벽과 창 두께가 추정되었다. 기사용 연료를 hot cell에 넣기 전에 차폐물질의 두께 추정을 위해 그 연료를 여러 시간 간격동안 저장용기 속에서 냉각시켰다는 가정을 했다. 여러 종류의 차폐물질이 고려되었으며 방사선원과 관측점과의 거리도 변화시켜 보았다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1997년도 추계학술대회논문집; 한국과학기술회관; 6 Nov. 1997
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• pp.177-183
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• 1997

The fuel rod for PWR is supported by the spring of the sapcer grid to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in the reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. Here considering the spring behaviour as a function of burnup the reaction forces of the springs are calculated by the finite element program developed herein to evaluate the integrity of the fuel rod from fretting. The results are compared with previous data and ANSYS for the validation of the program and procedures.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.173-179
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• 1998

The corrosion behavior of KOFA cladding which is a standard Zircaloy-4 manufactured by Westinghouse Specialty Metal Plant according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification was analyzed using the oxide measurement data of KOFA fuel irradiated in Kori-2 nuclear power plant. Analysis of the measured KOFA cladding oxidation showed that oxidation of KOFA cladding was lower than the design prediction based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Although the measured fuel rods have relatively low burnup and oxidation and the amount of the measured data are small, analysis of manufacturing and in-reactor operation conditions of KOFA cladding indicates that the differences in the manufacturing processes and chemical composition of the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 and KOFA cladding may have somewhat contributed to lower corrosion of KOFA cladding than the expected.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.67-73
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• 1997

By introducing the concept of overpressurization of rim pores due to dislocation punching, the total pressure exerted on the rim pores is estimated. Then this concept is combined with the assumption that all the fission gases produced in the rim region are retained in the rim region to calculate the rim porosity. Rim porosities calculated in this way are compared with measured data, which produces reasonable agreement. Finally a correlation for the thermal conductivity of the rim region is obtained using the hypothesis that the rim region consists of pores and fully dense material of UO$_2$.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1031-1043
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• 2017

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1339-1348
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• 2018

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.