• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4722-4730
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• 2022

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation - Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the depletion matrix are excluded in the corrector, and hence the corrector calculation involves only the reaction components, which can be efficiently solved with the Talyor Expansion Method (TEM). The numerical test shows that the new scheme substantially reduces computing time without loss of accuracy in comparison with the conventional scheme using CRAM (Chebyshev Rational Approximation Method), especially when the substep calculations are applied. The depletion calculation and source term estimation capability of BESNA are verified and validated through several problems, where results from BESNA are compared with those calculated by other codes as well as measured data. The analysis results show the computational efficiency of the new modified scheme and the reliability of BESNA in both isotopic predictions and source term estimations.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.650-659
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• 2016

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

• Nuclear Engineering and Technology
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• 제14권1호
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• pp.22-40
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• 1982

조화 단진동자 파동함수를 써서 원자핵의 LS에너지 행열요소를 계산하였다. 범위는 1$_1$= $l_{s}$ , $l_2$=lp, $l_3$=ld, 2s, $l_4$=1f, 2p, $l_{5}$ =1g, 2d, 3s라 ( $l_{i}$ $l_{i+1}$｜ $l_{i}$ $l_{i+1}$)의 배치에 대한 것이었다. 계산결과는 Talmi적분 $I_1$과 Slater 적분 $F^{k}$ 를 써서 표시하였다. 또 여러가지 합법칙을 유도하고 이를 써서 계산의 결과를 검산하였다.하였다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2095-2103
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• 2021

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.218-231
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• 2002

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high- level radioactive waste repository by adopting a finite difference method utilizing the control- volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical harriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three- dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of $^{234}$ Ulongrightarrow$^{230}$ Thlongrightarrow$^{226}$ Ra, which is the most important chain as far as the human environment is concerned.

• 열처리공학회지
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• 제18권5호
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• pp.288-296
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• 2005

The corrosion behaviors of Zr-based alloys were very sensitive to their microstructures which were determined by manufacturing process. The specimens of Zr-based alloy named as HANA-4 for nuclear fuel cladding were investigated in order to get the optimized manufacturing process such as the intermediate annealing temperature and cold working steps after the ${\beta}$ quenching. From the microstructural analysis, cold worked microstructure of the samples was changed to the recrystallized microstructure by performed process. The corrosion behaviors of HANA-4 alloy were affected by the different manufacturing process. The ${\beta}$-Zr phase was formed in the matrix and the Nb concentration in the ${\beta}$-Zr phase was increased as progressing the manufacturing process. So, it was found that the corrosion rate of HANA-4 alloy was affected by the Nb concentration in the matrix.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1345-1354
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• 2019

Clogging of sump strainers that filter the recirculation water in containment after a loss-of-coolant accident (LOCA) seriously impedes the continued cooling of nuclear reactor cores. In experiments examining the corrosion of aluminium alloy 6061, a common material in containment equipment, in borated solutions simulating the water chemistry of sump water after a LOCA, we found that Fe-bearing intermetallic particles, which were initially buried in the Al matrix, were progressively exposed as corrosion continued. Their cathodic nature $vis-{\grave{a}}-vis$ the Al matrix provoked continuous trenching around them until they were finally released into the test solution. Such particles released from Al alloy components in a reactor containment after a LOCA will be transported to the sump entrance with the recirculation flow and trapped by the debris bed that typically forms on the strainer surface, potentially aggravating strainer clogging. These Fe-bearing intermetallic particles, many of which had a rod or thin strip-like geometry, were identified to be mainly the cubic phase ${\alpha}_c-Al(Fe,Mn)Si$ with an average size of about $2.15{\mu}m$; 11.5 g of particles with a volume of about $3.2cm^3$ would be released with the dissolution of every 1 kg 6061 aluminium alloy.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.68-77
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• 1997

The reactor power control system is described in the fashion of the order increased LQR system. To obtain the optimal state feedback gain vectors, the weighting matrix of the performance function should be determined. Since the contentional method has some limitations, stochastic searching methods are investigated to optimize the LQR weighting matrix using the modified genetic algorithm combined with the simulated annealing, a new optimizing tool named the hybrid MGA-SA is developed to determine the weighting parameters of the LQR system. This optimizing tool provides a more systematic approach in designing the LQR system. Since it can be easily incorporated with any forms of the cost function, it also provides the great flexibility in the optimization problems.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1181-1188
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• 2017

This paper proposes the adoption of Monte Carlo perturbation theory to approximate the Jacobian matrix of coupled neutronics/thermal-hydraulics problems. The projected Jacobian is obtained from the eigenvalue decomposition of the fission matrix, and it is adopted to solve the coupled problem via the Newton method. This avoids numerical differentiations commonly adopted in Jacobian-free Newton-Krylov methods that tend to become expensive and inaccurate in the presence of Monte Carlo statistical errors in the residual. The proposed approach is presented and preliminarily demonstrated for a simple two-dimensional pressurized water reactor case study.