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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 46, Issue 6 - Dec 2014
Volume 46, Issue 5 - Oct 2014
Volume 46, Issue 4 - Aug 2014
Volume 46, Issue 3 - Jun 2014
Volume 46, Issue 2 - Apr 2014
Volume 46, Issue 1 - Feb 2014
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RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS
Chiba, Go ; Tsuji, Masashi ; Narabayashi, Tadashi ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 281~290
DOI : 10.5516/NET.01.2014.707
In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.
UNCERTAINTY PROPAGATION ANALYSIS FOR YONGGWANG NUCLEAR UNIT 4 BY MCCARD/MASTER CORE ANALYSIS SYSTEM
Park, Ho Jin ; Lee, Dong Hyuk ; Shim, Hyung Jin ; Kim, Chang Hyo ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 291~298
DOI : 10.5516/NET.01.2014.708
This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (
), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.
PROPAGATION OF NUCLEAR DATA UNCERTAINTIES FOR PWR CORE ANALYSIS
Cabellos, O. ; Castro, E. ; Ahnert, C. ; Holgado, C. ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 299~312
DOI : 10.5516/NET.01.2014.709
An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for
, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.
OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM) FOR LWRS - SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I)
Bratton, Ryan N. ; Avramova, M. ; Ivanov, K. ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 313~342
DOI : 10.5516/NET.01.2014.710
A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.:
radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of
NUCLEAR DATA UNCERTAINTY AND SENSITIVITY ANALYSIS WITH XSUSA FOR FUEL ASSEMBLY DEPLETION CALCULATIONS
Zwermann, W. ; Aures, A. ; Gallner, L. ; Hannstein, V. ; Krzykacz-Hausmann, B. ; Velkov, K. ; Martinez, J.S. ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 343~352
DOI : 10.5516/NET.01.2014.711
Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.
NUCLEAR DATA UNCERTAINTY PROPAGATION FOR A TYPICAL PWR FUEL ASSEMBLY WITH BURNUP
Rochman, D. ; Sciolla, C.M. ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 353~362
DOI : 10.5516/NET.01.2014.712
The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The "Fast Total Monte Carlo" method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on
, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.
PUMP DESIGN AND COMPUTATIONAL FLUID DYNAMIC ANALYSIS FOR HIGH TEMPERATURE SULFURIC ACID TRANSFER SYSTEM
Choi, Jung-Sik ; Shin, Young-Joon ; Lee, Ki-Young ; Yun, Yong-Sup ; Choi, Jae-Hyuk ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 363~372
DOI : 10.5516/NET.02.2014.011
In this study, we proposed a newly designed sulfuric acid transfer system for the sulfur-iodine (SI) thermochemical cycle. The proposed sulfuric acid transfer system was evaluated using a computational fluid dynamics (CFD) analysis for investigating thermodynamic/hydrodynamic characteristics and material properties. This analysis was conducted to obtain reliable continuous operation parameters; in particular, a thermal analysis was performed on the bellows box and bellows at amplitudes and various frequencies (0.1, 0.5, and 1.0 Hz). However, the high temperatures and strongly corrosive operating conditions of the current sulfuric acid system present challenges with respect to the structural materials of the transfer system. To resolve this issue, we designed a novel transfer system using polytetrafluoroethylene (PTFE,
) as a bellows material for the transfer of sulfuric acid. We also carried out a CFD analysis of the design. The CFD results indicated that the maximum applicable temperature of PTFE is about 533 K (
), even though its melting point is around 600 K. This result implies that the PTFE is a potential material for the sulfuric acid transfer system. The CFD simulations also confirmed that the sulfuric acid transfer system was designed properly for this particular investigation.
PREDICTION OF THE REACTOR VESSEL WATER LEVEL USING FUZZY NEURAL NETWORKS IN SEVERE ACCIDENT CIRCUMSTANCES OF NPPS
Park, Soon Ho ; Kim, Dae Seop ; Kim, Jae Hwan ; Na, Man Gyun ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 373~380
DOI : 10.5516/NET.04.2013.087
Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.
FUNCTIONAL VERIFICATION OF A SAFETY CLASS CONTROLLER FOR NPPS USING A UVM REGISTER MODEL
Kim, Kyuchull ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 381~386
DOI : 10.5516/NET.04.2013.080
A highly reliable safety class controller for NPPs (Nuclear Power Plants) is mandatory as even a minor malfunction can lead to disastrous consequences for people, the environment or the facility. In order to enhance the reliability of a safety class digital controller for NPPs, we employed a diversity approach, in which a PLC-type controller and a PLD-type controller are to be operated in parallel. We built and used structured testbenches based on the classes supported by UVM for functional verification of the PLD-type controller designed for NPPs. We incorporated a UVM register model into the testbenches in order to increase the controllability and the observability of the DUT(Device Under Test). With the increased testability, we could easily verify the datapaths between I/O ports and the register sets of the DUT, otherwise we had to perform black box tests for the datapaths, which is very cumbersome and time consuming. We were also able to perform constrained random verification very easily and systematically. From the study, we confirmed the various advantages of using the UVM register model in verification such as scalability, reusability and interoperability, and set some design guidelines for verification of the NPP controllers.
OPTIMIZATION OF OPERATION PARAMETERS OF 80-KEV ELECTRON GUN
Kim, Jeong Dong ; Lee, Yongdeok ; Kang, Heung Sik ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 387~394
DOI : 10.5516/NET.06.2013.092
A Slowing Down Time Spectrometer (SDTS) system is a highly efficient technique for isotopic nuclear material content analysis. SDTS technology has been used to analyze spent nuclear fuel and the pyro-processing of spent fuel. SDTS requires an external neutron source to induce the isotopic fissile fission. A high intensity neutron source is required to ensure a high for a good fissile fission. The electron linear accelerator system was selected to generate proper source neutrons efficiently. As a first step, the electron generator of an 80-keV electron gun was manufactured. In order to produce the high beam power from electron linear accelerator, a proper beam current is required form the electron generator. In this study, the beam current was measured by evaluating the performance of the electron generator. The beam current was determined by five parameters: high voltage at the electron gun, cathode voltage, pulse width, pulse amplitude, and bias voltage at the grid. From the experimental results under optimal conditions, the high voltage was determined to be 80 kV, the pulse width was 500 ns, and the cathode voltage was from 4.2 V to 4.6 V. The beam current was measured as 1.9 A at maximum. These results satisfy the beam current required for the operation of an electron linear accelerator.
SEPARATION OF CsCl FROM LiCl-CsCl MOLTEN SALT BY COLD FINGER MELT CRYSTALLIZATION
Versey, Joshua R. ; Phongikaroon, Supathorn ; Simpson, Michael F. ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 395~406
DOI : 10.5516/NET.06.2013.082
This study provides a fundamental understanding of a cold finger melt crystallization technique by exploring the heat and mass transfer processes of cold finger separation. A series of experiments were performed using a simplified LiCl-CsCl system by varying initial CsCl concentrations (1, 3, 5, and 7.5 wt%), cold finger cooling rates (7.4, 9.8, 12.3, and 14.9 L/min), and separation times (5, 10, 15, and 30 min). Results showed a potential recycling rate of 0.36 g/min with a purity of 0.33 wt% CsCl in LiCl. A CsCl concentrated drip formation was found to decrease crystal purity especially for smaller crystal formations. Dimensionless heat and mass transfer correlations showed that separation production is primarily influenced by convective transfer controlled by cooling gas flow rate, where correlations are more accurate for slower cooling gas flow rates.
THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR
Korkmaz, Mehmet E. ; Agar, Osman ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 407~412
DOI : 10.5516/NET.07.2013.050
In this research, we investigated the burnup characteristics and the conversion of fertile
in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning
fuel (fuel pin 1) and
fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.
EFFECTS OF TEMPERING AND PWHT ON MICROSTRUCTURES AND MECHANICAL PROPERTIES OF SA508 GR.4N STEEL
Lee, Ki-Hyoung ; Jhung, Myung Jo ; Kim, Min-Chul ; Lee, Bong-Sang ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 413~422
DOI : 10.5516/NET.07.2013.088
Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of
then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at
. The sample tempered at
with PWHT at
shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at
IN-PILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR
Kim, Hyun-Gil ; Park, Jeong-Yong ; Jeong, Yong-Hwan ; Koo, Yang-Hyun ; Yoo, Jong-Sung ; Mok, Yong-Kyoon ; Kim, Yoon-Ho ; Suh, Jung-Min ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 423~430
DOI : 10.5516/NET.07.2013.093
An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.
EVALUATION OF PH CONTROL AGENTS INFLUENCING ON CORROSION OF CARBON STEEL IN SECONDARY WATER CHEMISTRY CONDITION OF PRESSURIZED WATER REACTOR
Rhee, In Hyoung ; Jung, Hyunjun ; Cho, Daechul ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 431~438
DOI : 10.5516/NET.09.2013.076
The effect of various pH agents on the corrosion behavior of carbon steel was investigated under a simulated secondary water chemistry condition of a pressurized water reactor (PWR) in a laboratory, and the steel's corrosion performance was compared with the field data obtained from Uljin NPP unit 2 reactor. All tests were carried out at temperatures of
and pH of 8.5 - 10. The pH at a given temperature was controlled by adding different agents. Laboratory data indicate that the corrosion rate of carbon steel decreased as the pH increased under the test conditions and the highest corrosion rate was measured at
. This high corrosion rate may be related to high dissolution and instability of Fe oxide (
. It was also found that an addition of ethanolamine (ETA) to ammonia was more effectivefor anticorrosion than ammonia alone, and that mixed treatment reduced 50% of iron or more at pHs of 9.5 or higher, especially in the steam generator (SG) and the moisture separator & re-heater (MSR).
DEVELOPMENT OF AN AMPHIBIOUS ROBOT FOR VISUAL INSPECTION OF APR1400 NPP IRWST STRAINER ASSEMBLY
Jang, You Hyun ; Kim, Jong Seog ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 439~446
DOI : 10.5516/NET.09.2013.085
An amphibious inspection robot system (hereafter AIROS) is being developed to visually inspect the in-containment refueling storage water tank (hereafter IRWST) strainer in APR1400 instead of a human diver. Four IRWST strainers are located in the IRWST, which is filled with boric acid water. Each strainer has 108 sub-assembly strainer fin modules that should be inspected with the VT-3 method according to Reg. guide 1.82 and the operation manual. AIROS has 6 thrusters for submarine voyage and 4 legs for walking on the top of the strainer. An inverse kinematic algorithm was implemented in the robot controller for exact walking on the top of the IRWST strainer. The IRWST strainer has several top cross braces that are extruded on the top of the strainer, which can be obstacles of walking on the strainer, to maintain the frame of the strainer. Therefore, a robot leg should arrive at the position beside the top cross brace. For this reason, we used an image processing technique to find the top cross brace in the sole camera image. The sole camera image is processed to find the existence of the top cross brace using the cross edge detection algorithm in real time. A 5-DOF robot arm that has multiple camera modules for simultaneous inspection of both sides can penetrate narrow gaps. For intuitive presentation of inspection results and for management of inspection data, inspection images are stored in the control PC with camera angles and positions to synthesize and merge the images. The synthesized images are then mapped in a 3D CAD model of the IRWST strainer with the location information. An IRWST strainer mock-up was fabricated to teach the robot arm scanning and gaiting. It is important to arrive at the designated position for inserting the robot arm into all of the gaps. Exact position control without anchor under the water is not easy. Therefore, we designed the multi leg robot for the role of anchoring and positioning. Quadruped robot design of installing sole cameras was a new approach for the exact and stable position control on the IRWST strainer, unlike a traditional robot for underwater facility inspection. The developed robot will be practically used to enhance the efficiency and reliability of the inspection of nuclear power plant components.
EDUCATIONAL EFFECTS OF RADIATION WORK-STUDY ACTIVITIES FOR ELEMENTARY, MIDDLE, AND HIGH SCHOOL STUDENTS
Han, Eun Ok ; Kim, Jae Rok ; Choi, Yoon Seok ;
Nuclear Engineering and Technology, volume 46, issue 3, 2014, Pages 447~460
DOI : 10.5516/NET.10.2013.091
The results of this study, suggest public communication to promote the use of radiation as follows: first, suitable information for the recipient's perception patterns should be provided, as there is a difference in risk perception and acceptance between the experts and the public. Thus, information on the necessity of nuclear power should be provided to the public, while information based on technical risks is provided by the experts. Second, since the levels of perception, knowledge, and attitudes increased highly for sectors which use radiation after the class, classes should be provided continuously to increase students' perception, knowledge, and attitude, which are all preemptive variables which induce positive behavioral changes. Third, since the seven sectors which use radiation are highly correlated, arguments for the necessity of other sectors should be based on the necessity of the medical sector.