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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 43, Issue 6 - Dec 2011
Volume 43, Issue 5 - Oct 2011
Volume 43, Issue 4 - Aug 2011
Volume 43, Issue 3 - Jun 2011
Volume 43, Issue 2 - Apr 2011
Volume 43, Issue 1 - Feb 2011
Selecting the target year
MULTISCALE MODELLING FOR THE FISSION GAS BEHAVIOUR IN THE TRANSURANUS CODE
Van Uffelen, P. ; Pastore, G. ; Di Marcello, V. ; Luzzi, L. ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 477~488
DOI : 10.5516/NET.2011.43.6.477
A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of
under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.
DEVELOPMENT OF THE ENIGMA FUEL PERFORMANCE CODE FOR WHOLE CORE ANALYSIS AND DRY STORAGE ASSESSMENTS
Rossiter, Glyn ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 489~498
DOI : 10.5516/NET.2011.43.6.489
UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.
FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO
AND MOX FUEL
Lee, Byung-Ho ; Koo, Yang-Hyun ; Oh, Jae-Yong ; Cheon, Jin-Sik ; Tahk, Young-Wook ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 499~508
DOI : 10.5516/NET.2011.43.6.499
The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in
fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the
irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and
fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.
RECENT UPDATES TO NRC FUEL PERFORMANCE CODES AND PLANS FOR FUTURE IMPROVEMENTS
Geelhood, Kenneth ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 509~522
DOI : 10.5516/NET.2011.43.6.509
FRAPCON-3.4a and FRAPTRAN 1.4 are the most recent versions of the U.S. Nuclear Regulatory Commission (NRC) steady-state and transient fuel performance codes, respectively. These codes have been assessed against separate effects data and integral assessment data and have been determined to provide a best estimate calculation of fuel performance. Recent updates included in FRAPCON-3.4a include updated material properties models, models for new fuel and cladding types, cladding finite element analysis capability, and capability to perform uncertainty analyses and calculate upper tolerance limits for important outputs. Recent updates included in FRAPTRAN 1.4 include: material properties models that are consistent with FRAPCON-3.4a, cladding failure models that are applicable for loss-of coolant-accident and reactivity initiated accident modeling, and updated heat transfer models. This paper briefly describes these code updates and data assessments, highlighting the particularly important improvements and data assessments. This paper also discusses areas of improvements that will be addressed in upcoming code versions.
FURTHER EVALUATION OF A STOCHASTIC MODEL APPLIED TO MONOENERGETIC SPACE-TIME NUCLEAR REACTOR KINETICS
Ha, Pham Nhu Viet ; Kim, Jong-Kyung ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 523~530
DOI : 10.5516/NET.2011.43.6.523
In a previous study, the stochastic space-dependent kinetics model (SSKM) based on the forward stochastic model in stochastic kinetics theory and the Ito stochastic differential equations was proposed for treating monoenergetic space-time nuclear reactor kinetics in one dimension. The SSKM was tested against analog Monte Carlo calculations, however, for exemplary cases of homogeneous slab reactors with only one delayed-neutron precursor group. In this paper, the SSKM is improved and evaluated with more realistic and complicated cases regarding several delayed-neutron precursor groups and heterogeneous slab reactors in which the extraneous source or reactivity can be introduced locally. Furthermore, the source level and the initial conditions will also be adjusted to investigate the trends in the variances of the neutron population and fission product levels across the reactor. The results indicate that the improved SSKM is in good agreement with the Monte Carlo method and show how the variances in population dynamics can be controlled.
APPLICATION OF BACKWARD DIFFERENTIATION FORMULA TO SPATIAL REACTOR KINETICS CALCULATION WITH ADAPTIVE TIME STEP CONTROL
Shim, Cheon-Bo ; Jung, Yeon-Sang ; Yoon, Joo-Il ; Joo, Han-Gyu ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 531~546
DOI : 10.5516/NET.2011.43.6.531
The backward differentiation formula (BDF) method is applied to a three-dimensional reactor kinetics calculation for efficient yet accurate transient analysis with adaptive time step control. The coarse mesh finite difference (CMFD) formulation is used for an efficient implementation of the BDF method that does not require excessive memory to store old information from previous time steps. An iterative scheme to update the nodal coupling coefficients through higher order local nodal solutions is established in order to make it possible to store only node average fluxes of the previous five time points. An adaptive time step control method is derived using two order solutions, the fifth and the fourth order BDF solutions, which provide an estimate of the solution error at the current time point. The performance of the BDF- and CMFD-based spatial kinetics calculation and the adaptive time step control scheme is examined with the NEACRP control rod ejection and rod withdrawal benchmark problems. The accuracy is first assessed by comparing the BDF-based results with those of the Crank-Nicholson method with an exponential transform. The effectiveness of the adaptive time step control is then assessed in terms of the possible computing time reduction in producing sufficiently accurate solutions that meet the desired solution fidelity.
ANALYSIS OF EQUILIBRIUM METHODS FOR THE COMPUTATIONAL MODEL OF THE MARK-IV ELECTR OREFINER
Cumberland, Riley ; Hoover, Robert ; Phongikaroon, Supathorn ; Yim, Man-Sung ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 547~556
DOI : 10.5516/NET.2011.43.6.547
Two computational methods for determining equilibrium states for the Mark-IV electrorefiner (ER) have been assessed to improve the current computational electrorefiner model developed at University of Idaho. Both methods were validated against measured data to better understand their effects on the calculation of the equilibrium compositions in the ER. In addition, a sensitivity study was performed on the effect of specific unknown activity coefficients-including sodium in molten cadmium, zirconium in molten cadmium, and sodium chloride in molten LiCl-KCl. Both computational methods produced identical results, which stayed within the 95% confidence interval of the experimental data. Furthermore, sensitivity to unavailable activity coefficients was found to be low (a change in concentration of less than 3 ppm).
APPLICATION OF COLD SPRAY COATING TECHNIQUE TO AN UNDERGROUND DISPOSAL COPPER CANISTER AND ITS CORROSION PROPERTIES
Lee, Min-Soo ; Choi, Heui-Joo ; Choi, Jong-Won ; Kim, Hyung-Jun ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 557~566
DOI : 10.5516/NET.2011.43.6.557
A cold spray coating (CSC) of copper was studied for its application to a high-level radioactive waste (HLW) disposal canister. Several copper coatings of 10 mm thick were fabricated using two kinds of copper powders with different oxygen contents, and SS 304 and nodular cast iron were used as their base metal substrates. The fabricated CSC coppers showed a high tensile strength but were brittle in comparison with conventional non-coating copper, hereinafter defined to as "commercial copper". The corrosion behavior of CSC coppers was evaluated by comparison with commercial coppers, such as extruded and forged coppers. The polarization test results showed that the corrosion potential of the CSC coppers was closely related to its purity; low-purity (i.e., high oxygen content) copper exhibited a lower corrosion potential, and high-purity copper exhibited a relatively high corrosion potential. The corrosion rate converted from the measured corrosion current was not, however, dependent on its purity: CSC copper showed a little higher rate than that of commercial copper. Immersion tests in aqueous HCl solution showed that CSC coppers were more susceptible to corrosion, i.e., they had a higher corrosion rate. However, the difference was not significant between commercial copper and high-purity CSC copper. The decrease of corrosion was observed in a humid air test presumably due to the formation of a protective passive film. In conclusion, the results of this study indicate that CSC application of copper could be a useful option for fabricating a copper HLW disposal canister.
SENSITIVITY ANALYSIS TO EVALUATE THE TRANSPORT PROPERTIES OF CdZnTe DETECTORS USING ALPHA PARTICLES AND LOW-ENERGY GAMMA-RAYS
Kim, Kyung-O ; Ahn, Woo-Sang ; Kwon, Tae-Je ; Kim, Soon-Young ; Kim, Jong-Kyung ; Ha, Jang-Ho ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 567~572
DOI : 10.5516/NET.2011.43.6.567
A sensitivity analysis of the methods used to evaluate the transport properties of a CdZnTe detector was performed using two different radiations (
particle and gamma-ray) emitted from an
source. The mobility-lifetime products of the electron-hole pair in a planar CZT detector (
) were determined by fitting the peak position as a function of biased voltage data to the Hecht equation. To verify the accuracy of these products derived from
particles and low-energy gamma-rays, an energy spectrum considering the transport property of the CZT detector was simulated through a combination of the deposited energy and the charge collection efficiency at a specific position. It was found that the shaping time of the amplifier module significantly affects the determination of the (
) products; the
particle method was stabilized with an increase in the shaping time and was less sensitive to this change compared to when the gamma-ray method was used. In the case of the simulated energy spectrum with transport properties evaluated by the
particle method, the peak position and tail were slightly different from the measured result, whereas the energy spectrum derived from the low-energy gamma-ray was in good agreement with the experimental results. From these results, it was confirmed that low-energy gamma-rays are more useful when seeking to obtain the transport properties of carriers than
particles because the methods that use gamma-rays are less influenced by the surface condition of the CZT detector. Furthermore, the analysis system employed in this study, which was configured by a combination of Monte Carlo simulation and the Hecht model, is expected to be highly applicable to the study of the characteristics of CZT detectors.
A PROCEDURE FOR GENERATING IN-CABINET RESPONSE SPECTRA BASED ON STATE-SPACE MODEL IDENTIFICATION BY IMPACT TESTING
Cho, Sung-Gook ; Cui, Jintao ; Kim, Doo-Kie ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 573~582
DOI : 10.5516/NET.2011.43.6.573
The in-cabinet response spectrum is used to define the input motion in the seismic qualification of instruments and devices mounted inside an electrical cabinet. This paper presents a procedure for generating the in-cabinet response spectrum for electrical equipment based on in-situ testing by an impact hammer. The proposed procedure includes an algorithm to build the relationship between the impact forces and the measured acceleration responses of cabinet structures by estimating the state-space model. This model is used to predict seismic responses to the equivalent earthquake forces. Three types of structural model are analyzed for numerical verification of the proposed method. A comparison of predicted and simulated response spectra shows good convergence, demonstrating the potential of the proposed method to predict the response spectra for real cabinet structures using vibration tests. The presented procedure eliminates the uncertainty associated with constructing an analytical model of the electrical cabinet, which has complex mass distribution and stiffness.
RADIOLOGICAL CHARACTERISTICS OF DECOMMISSIONING WASTE FROM A CANDU REACTOR
Cho, Dong-Keun ; Choi, Heui-Joo ; Ahmed, Rizwan ; Heo, Gyun-Young ;
Nuclear Engineering and Technology, volume 43, issue 6, 2011, Pages 583~592
DOI : 10.5516/NET.2011.43.6.583
The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be
, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.