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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 42, Issue 6 - Dec 2010
Volume 42, Issue 5 - Oct 2010
Volume 42, Issue 4 - Aug 2010
Volume 42, Issue 3 - Jun 2010
Volume 42, Issue 2 - Apr 2010
Volume 42, Issue 1 - Feb 2010
Selecting the target year
ASSESSMENT OF CFD CODES USED IN NUCLEAR REACTOR SAFETY SIMULATIONS
Smith, Brian L. ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 339~364
DOI : 10.5516/NET.2010.42.4.339
Following a joint OECD/NEA-IAEA-sponsored meeting to define the current role and future perspectives of the application of Computational Fluid Dynamics (CFD) to nuclear reactor safety problems, three Writing Groups were created, under the auspices of the NEA working group WGAMA, to produce state-of-the-art reports on different aspects of the subject. The work of the second group, WG2, was to document the existing assessment databases for CFD simulation in the context of Nuclear Reactor Safety (NRS) analysis, to gain a measure of the degree of quality and trust in CFD as a numerical analysis tool, and to take initiatives to extend the existing databases. The group worked over the period of 2003-2007 and produced a final state-of-the-art report. The present paper summarises the material gathered during the study, illustrating the points with a few highlights. A total of 22 safety issues were identified for which the application of CFD was considered to potentially bring real benefits in terms of better understanding and increased safety. A list of the existing databases was drawn up and synthesised, both from the nuclear area and from other parallel, non-nuclear, industrial activities. The gaps in the technology base were also identified and discussed. In order to initiate new ways of bringing experimentalists and numerical analysts together, an international workshop -- CFD4NRS (the first in a series) -- was organised, a new blind benchmark activity was set up based on turbulent mixing in T-junctions, and a Wiki-type web portal was created to offer online access to the material put together by the group giving the reader the opportunity to update and extend the contents to keep the information source topical and dynamic.
EXTENSION OF CFD CODES APPLICATION TO TWO-PHASE FLOW SAFETY PROBLEMS
Bestion, Dominique ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 365~376
DOI : 10.5516/NET.2010.42.4.365
This paper summarizes the results of a Writing Group on the Extension of CFD codes to two-phase flow safety problems, which was created by the Group for Analysis and Management of Accidents of the Nuclear Energy Agency' Committee on the Safety of Nuclear Installations (NEA-CSNI). Two-phase CFD used for safety investigations may predict small scale flow processes, which are not seen by system thermalhydraulic codes. However, the two-phase CFD models are not as mature as those in the single phase CFD and potential users need some guidance for proper application. In this paper, a classification of various modelling approaches is proposed. Then, a general multi-step methodology for using two-phase-CFD is explained, including a preliminary identification of flow processes, a model selection, and a verification and validation process. A list of 26 nuclear reactor safety issues that could benefit from investigations at the CFD scale is identified. Then, a few issues are analyzed in more detail, and a preliminary state-of-the-art is proposed and the remaining gaps in the existing approaches are identified. Finally, guidelines for users are proposed.
DEVELOPMENT OF BEST PRACTICE GUIDELINES FOR CFD IN NUCLEAR REACTOR SAFETY
Mahaffy, John ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 377~381
DOI : 10.5516/NET.2010.42.4.377
In 2007 the Nuclear Energy Agency's Committee on the Safety of Nuclear Installations published Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety. This paper provides an overview of the document' contents and highlights a few of its recommendations. The document covers the full extent of a CFD analysis from initial problem definition and selection of an appropriate tool for the analysis, through final documentation of results. It provides advice on selection of appropriate simulation software, mesh construction, and selection of physical models. In addition it contains extensive discussion of the verification and validation process that should accompany any high-quality CFD analysis.
CFD ANALYSIS OF TURBULENT JET BEHAVIOR INDUCED BY A STEAM JET DISCHARGED THROUGH A VERTICAL UPWARD SINGLE HOLE IN A SUBCOOLED WATER POOL
Kang, Hyung-Seok ; Song, Chul-Hwa ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 382~393
DOI : 10.5516/NET.2010.42.4.382
Thermal mixing by steam jets in a pool is dominantly influenced by a turbulent water jet generated by the condensing steam jets, and the proper prediction of this turbulent jet behavior is critical for the pool mixing analysis. A turbulent jet flow induced by a steam jet discharged through a vertical upward single hole into a subcooled water pool was subjected to computational fluid dynamics (CFD) analysis. Based on the small-scale test data derived under a horizontal steam discharging condition, this analysis was performed to validate a CFD method of analysis previously developed for condensing jet-induced pool mixing phenomena. In previous validation work, the CFD results and the test data for a limited range of radial and axial directions were compared in terms of profiles of the turbulent jet velocity and temperature. Furthermore, the behavior of the turbulent jet induced by the steam jet through a horizontal single hole in a subcooled water pool failed to show the exact axisymmetric flow pattern with regards to an overall pool mixing, whereas the CFD analysis was done with an axisymmetric grid model. Therefore, another new small-scale test was conducted under a vertical upward steam discharging condition. The purpose of this test was to generate the velocity and temperature profiles of the turbulent jet by expanding the measurement ranges from the jet center to a location at about 5% of
and 10 cm to 30 cm from the exit of the discharge nozzle. The results of the new CFD analysis show that the recommended CFD model of the high turbulent intensity of 40% for the turbulent jet and the fine mesh grid model can accurately predict the test results within an error rate of about 10%. In this work, the turbulent jet model, which is used to simply predict the temperature and velocity profiles along the axial and radial directions by means of the empirical correlations and Tollmien's theory was improved on the basis of the new test data. The results validate the CFD model of analysis. Furthermore, the turbulent jet model developed in this study can be used to analyze pool thermal mixing when an ellipsoidal steam jet is discharged under a high steam mass flux in a subcooled water pool.
KEY IMPACT PARAMETERS FOR APPLICATION OF ALTERNATIVE SOURCE TERM TO KORI UNIT 1
Lee, Seung-Chan ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 394~413
DOI : 10.5516/NET.2010.42.4.394
The object of this paper is to identify the key elements that impact a radiation dose at EAB (Exclusion Area Boundary). This study is based on the AST (Alternative Source Terms) as defined in Regulatory Guide 1.183. The LOCA (Loss of Coolant Accident) and the LRA (Locked Rotor Accident) are selected as limiting cases. A sensitivity analysis of accidental behavior with respect to various parameters during LOCA and LRA at Kori Unit 1 is also undertaken for the following objectives: to determine the limiting parameters, to find the impact trend of the radiation dose, and to find the safety margin between AST and TID (Technical Information Document) methodologies. This work confirms that key parameters are particulate removal rate, decontamination factor, iodine chemical form, gap fraction, partitioning factor, and the impact of isotopes group. Comparing TID with AST, the radiation dose of TID is about 80% greater than that of AST under a LOCA, and about 60% greater than that of AST for the case of a LRA; thus the safety margin is remarkably increased when the AST is used. In this work, the sensitivity analysis results are presented in terms of a sensitivity index called the "NDD (Normalized Dose Difference)", which compares the impact of parameters with that of a reference case. These values are derived by using a combination of the leak rate (primary to secondary), iodine chemical form, gap fraction, partitioning factor, spray removal rate, source term, and other variables.
OPTIMIZATION OF THE TEST INTERVALS OF A NUCLEAR SAFETY SYSTEM BY GENETIC ALGORITHMS, SOLUTION CLUSTERING AND FUZZY PREFERENCE ASSIGNMENT
Zio, E. ; Bazzo, R. ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 414~425
DOI : 10.5516/NET.2010.42.4.414
In this paper, a procedure is developed for identifying a number of representative solutions manageable for decision-making in a multiobjective optimization problem concerning the test intervals of the components of a safety system of a nuclear power plant. Pareto Front solutions are identified by a genetic algorithm and then clustered by subtractive clustering into "families". On the basis of the decision maker's preferences, each family is then synthetically represented by a "head of the family" solution. This is done by introducing a scoring system that ranks the solutions with respect to the different objectives: a fuzzy preference assignment is employed to this purpose. Level Diagrams are then used to represent, analyze and interpret the Pareto Fronts reduced to the head-of-the-family solutions.
MATERIAL INVESTIGATION AND ANALYSIS USING CHARACTERISTIC X-RAY
Oh, Gyu-Bum ; Lee, Won-Ho ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 426~433
DOI : 10.5516/NET.2010.42.4.426
The characteristic X-rays emitted from materials after gamma ray exposure was simulated and measured. A CdTe semiconductor detector and a
radiation source were used for energy spectroscopy. The types of materials could be identified by comparing the measured energy spectrum with the theoretical X-ray transition energy of the material. The sample composition was represented by the
-line (Siegbahn notations), which has the highest intensity among the characteristic X-rays of each atom. The difference between the theoretic prediction and the experimental result of K-line measurement was < 0.61% even if the characteristic X-rays from several materials were measured simultaneously. 2D images of the mixed materials were acquired with very high selectivity.
ESTIMATION OF ALUMINUM AND ARGON ACTIVATION SOURCES IN THE HANARO COOLANT
Jun, Byung-Jin ; Lee, Byung-Chul ; Kim, Myung-Seop ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 434~441
DOI : 10.5516/NET.2010.42.4.434
The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.
NEW EVALUATION METHODS FOR RADIAL UNIFORMITY IN NEUTRON TRANSMUTATION DOPING
Kim, Hak-Sung ; Lim, Jae-Yong ; Pyeon, Cheol-Ho ; Misawa, Tsuyoshi ; Shiroya, Seiji ; Park, Sang-Jun ; Kim, Myong-Seop ; Oh, Soo-Youl ; Jun, Byung-Jin ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 442~449
DOI : 10.5516/NET.2010.42.4.442
Recently, the neutron irradiation for large diameter silicon (Si)-ingots of more than 8" diameter is requested to satisfy the demand for the neutron transmutation doping silicon (NTD-Si). By increasing the Si-ingot diameter, the radial non-uniformity becomes larger due to the neutron attenuation effect, which results in a limit of the feasible diameter of the Si-ingot. The current evaluation method has a certain limit to precisely evaluate the radial uniformity of Si-ingot because the current evaluation method does not consider the effect of the Si-ingot diameter on the radial uniformity. The objective of this study is to propose a new evaluation method of radial uniformity by improving the conventional evaluation approach. To precisely predict the radial uniformity of a Si-ingot with large diameter, numerical verification is conducted through comparison with the measured data and introducing the new evaluation method. A new concept of a gradient is introduced as an alternative approach of radial uniformity evaluation instead of the radial resistivity gradient (RRG) interpretation. Using the new concept of gradient, the normalized reaction rate gradient (NRG) and the surface normalized reaction rate gradient (SNRG) are described. By introducing NRG, the radial uniformity can be evaluated with one certain standard regardless of the ingot diameter and irradiation condition. Furthermore, by introducing SNRG, the uniformity on the Si-ingot surface, which is ignored by RRG and NRG, can be evaluated successfully. Finally, the radial uniformity flattening methods are installed by the stainless steel thermal neutron filter and additional Si-pipe to reduce SNRG.
COLLAPSE PRESSURE ESTIMATES AND THE APPLICATION OF A PARTIAL SAFETY FACTOR TO CYLINDERS SUBJECTED TO EXTERNAL PRESSURE
Yoo, Yeon-Sik ; Huh, Nam-Su ; Choi, Suhn ; Kim, Tae-Wan ; Kim, Jong-In ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 450~459
DOI : 10.5516/NET.2010.42.4.450
The present paper investigates the collapse pressure of cylinders with intermediate thickness subjected to external pressure based on detailed elastic-plastic finite element (FE) analyses. The effect of the initial ovality of the tube on the collapse pressure was explicitly considered in the FE analyses. Based on the present FE results, the analytical yield locus, considering the interaction between the plastic collapse and local instability due to initial ovality, was also proposed. The collapse pressure values based on the proposed yield locus agree well with the present FE results; thus, the validity of the proposed yield locus for the thickness range of interest was verified. Moreover, the partial safety factor concept based on the structural reliability theory was also applied to the proposed collapse pressure estimation model, and, thus, the priority of importance of respective parameter constituting for the collapse of cylinders under external pressure was estimated in this study. From the application of the partial safety factor concept, the yield strength was concluded to be the most sensitive, and the initial ovality of tube was not so effective in the proposed collapse pressure estimation model. The present deterministic and probabilistic results are expected to be utilized in the design and maintenance of cylinders subjected to external pressure with initial ovality, such as the once-through type steam generator.
OPTIMIZATION OF THE PARAMETERS OF FEEDWATER CONTROL SYSTEM FOR OPR1000 NUCLEAR POWER PLANTS
Kim, Ung-Soo ; Song, In-Ho ; Sohn, Jong-Joo ; Kim, Eun-Kee ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 460~467
DOI : 10.5516/NET.2010.42.4.460
In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.
AN INVESTIGATION INTO RADIATION LEVELS ASSOCIATED WITH DISMANTLING THE KOREA RESEARCH REACTOR
Choi, Geun-Sik ; Kim, Hee-Reyoung ; Han, Moon-Hee ;
Nuclear Engineering and Technology, volume 42, issue 4, 2010, Pages 468~473
DOI : 10.5516/NET.2010.42.4.468
We confirmed that the dismantling of two research reactors with thermal power of
, respectively, reveals no significant difference between the radiation levels of the research reactor site and the surrounding environment far away from it, from the radiation level aspect. Radiation dose and radioactivity were measured at monitoring points around the research reactor site of the Korea Atomic Energy Research Institute (KAERI) in Seoul and comparison points 0.5 km to 3.3 km from the site. To grasp trends in the radiation levels during dismantling from the end of 2002 to the end of 2007, the gamma radiation dose rate, the accumulated dose, and the radioactivity of the strontium, tritium, and gamma isotopes were statistically treated and estimated. The averages of these items between the two groups, the research reactor site and comparison points, were assessed by applying a T-test with a significance level of 0.05. P-values found by using the T-test were from 0.12 to 0.83 where the values were much higher than the significance level. As a result, no difference was observed between the radiation levels at the research reactor site and at the comparison points by this T-test. This study showed that dismantling activity of the Korea Research Reactor of the Seoul site did not expose the public or the environment to harmful levels of radiation.