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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 44, Issue 8 - Dec 2012
Volume 44, Issue 7 - Oct 2012
Volume 44, Issue 6 - Aug 2012
Volume 44, Issue 5 - Jun 2012
Volume 44, Issue 4 - May 2012
Volume 44, Issue 3 - Apr 2012
Volume 44, Issue 2 - Mar 2012
Volume 44, Issue 1 - Feb 2012
Selecting the target year
ANALYSIS OF THE ISP-50 DIRECT VESSEL INJECTION SBLOCA IN THE ATLAS FACILITY WITH THE RELAP5/MOD3.3 CODE
Sharabi, Medhat ; Freixa, Jordi ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 709~718
DOI : 10.5516/NET.02.2012.712
The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.
TRACE V5 CODE APPLICATION DVI LINE BREAK LOCA USING ATLAS FACILITY
Veronese, Fabio ; Kozlowsk, Tomasz ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 719~726
DOI : 10.5516/NET.02.2012.713
The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.
Kwon, Tae-Soon ; Lee, S.T. ; Euh, D.J. ; Chu, I.C. ; Youn, Y.J. ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 727~734
DOI : 10.5516/NET.02.2012.714
A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.
A FLOW AND PRESSURE DISTRIBUTION OF APR+ REACTOR UNDER THE 4-PUMP RUNNING CONDITIONS WITH A BALANCED FLOW RATE
Euh, D.J. ; Kim, K.H. ; Youn, Y.J. ; Bae, J.H. ; Chu, I.C. ; Kim, J.T. ; Kang, H.S. ; Choi, H.S. ; Lee, S.T. ; Kwon, T.S. ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 735~744
DOI : 10.5516/NET.02.2012.715
In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP (
ore Flow &
ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.
A STUDY ON METHODOLOGY FOR IDENTIFYING CORRELATIONS BETWEEN LERF AND EARLY FATALITY
Kang, Kyungmin ; Jae, Moosung ; Ahn, Kwang-Il ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 745~754
DOI : 10.5516/NET.03.2011.036
The correlations between Large Early Release Frequency (LERF) and Early Fatality need to be investigated for risk-informed application and regulation. In Regulatory Guide (RG) -1.174, while there are decision-making criteria using the measures of Core Damage Frequency (CDF) and LERF, there are no specific criteria on LERF. Since there are both huge uncertainties and large costs needed in off-site consequence calculation, a LERF assessment methodology needs to be developed, and its correlation factor needs to be identified, for risk-informed decision-making. A new method for estimating off-site consequence has been presented and performed for assessing health effects caused by radioisotopes released from severe accidents of nuclear power plants in this study. The MACCS2 code is used for validating the source term quantitatively regarding health effects, depending on the release characteristics of radioisotopes during severe accidents. This study developed a method for identifying correlations between LERF and Early Fatality and validates the results of the model using the MACCS2 code. The results of this study may contribute to defining LERF and finding a measure for risk-informed regulations and risk-informed decision-making.
NEW RESULTS TO BDD TRUNCATION METHOD FOR EFFICIENT TOP EVENT PROBABILITY CALCULATION
Mo, Yuchang ; Zhong, Farong ; Zhao, Xiangfu ; Yang, Quansheng ; Cui, Gang ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 755~766
DOI : 10.5516/NET.03.2011.058
A Binary Decision Diagram (BDD) is a graph-based data structure that calculates an exact top event probability (TEP). It has been a very difficult task to develop an efficient BDD algorithm that can solve a large problem since its memory consumption is very high. Recently, in order to solve a large reliability problem within limited computational resources, Jung presented an efficient method to maintain a small BDD size by a BDD truncation during a BDD calculation. In this paper, it is first identified that Jung's BDD truncation algorithm can be improved for a more practical use. Then, a more efficient truncation algorithm is proposed in this paper, which can generate truncated BDD with smaller size and approximate TEP with smaller truncation error. Empirical results showed this new algorithm uses slightly less running time and slightly more storage usage than Jung's algorithm. It was also found, that designing a truncation algorithm with ideal features for every possible fault tree is very difficult, if not impossible. The so-called ideal features of this paper would be that with the decrease of truncation limits, the size of truncated BDD converges to the size of exact BDD, but should never be larger than exact BDD.
SELECTIVE REDUCTION OF ACTIVE METAL CHLORIDES FROM MOLTEN LiCl-KCl USING LITHIUM DRAWDOWN
Simpson, Michael F. ; Yoo, Tae-Sic ; Labrier, Daniel ; Lineberry, Michael ; Shaltry, Michael ; Phongikaroon, Supathorn ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 767~772
DOI : 10.5516/NET.06.2011.010
In support of optimizing electrorefining technology for treating spent nuclear fuel, lithium drawdown has been investigated for separating actinides from molten salt electrolyte. Drawdown reaction selectivity is a major issue that requires investigation, since the goal is to remove actinides while leaving the fission products and other components in the salt. A series of lithium drawdown tests with surrogate fission product chlorides was run to obtain selectivity data with non-radioactive salts, develop a predictive model, and draw conclusions about the viability of using this process with actinide-loaded salt. Results of tests with CsCl,
are reported here. Equilibrium was typically achieved in less than 10 hours of contact between lithium metal and molten salt under well-stirred conditions. Maintaining low oxygen and water impurity concentrations (<10 ppm) in the atmosphere was observed to be critical to minimize side reactions and maintain stable salt compositions. An equilibrium model has been formulated and fit to the experimental data. Good fits to the data were achieved. Based on analysis and results obtained to date, it is concluded that clean separation between minor actinides and lanthanides will be difficult to achieve using lithium drawdown.
EVALUATION OF GALVANIC CORROSION BEHAVIOR OF SA-508 LOW ALLOY STEEL AND TYPE 309L STAINLESS STEEL CLADDING OF REACTOR PRESSURE VESSEL UNDER SIMULATED PRIMARY WATER ENVIRONMENT
Kim, Sung-Woo ; Kim, Dong-Jin ; Kim, Hong-Pyo ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 773~780
DOI : 10.5516/NET.07.2011.054
The article presented is concerned with an evaluation of the corrosion behavior of SA-508 low alloy steel (LAS) and Type 309L stainless steel (SS) cladding of a reactor pressure vessel under the simulated primary water chemistry of a pressurized water reactor (PWR). The uniform corrosion and galvanic corrosion rates of SA-508 LAS and Type 309L SS were measured in three different control conditions: power operation, shutdown, and power operation followed by shutdown. In all conditions, the dissimilar metal coupling of SA-508 LAS and Type 309L SS exhibited higher corrosion rates than the SA-508 base metal itself due to severe galvanic corrosion near the cladding interface, while the corrosion of Type 309L in the primary water environment was minimal. The galvanic corrosion rate of the SA-508 LAS and Type 309L SS couple measured under the simulated power operation condition was much lower than that measured in the simulated shutdown condition due to the formation of magnetite on the metal surface in a reducing environment. Based on the experimental results, the corrosion rate of SA-508 LAS clad with Type 309L SS was estimated as a function of operating cycle simulated for a typical PWR.
NEW DEVELOPMENT OF HYPERGAM AND ITS TEST OF PERFORMANCE FOR γ-RAY SPECTRUM ANALYSIS
Park, B.G. ; Choi, H.D. ; Park, C.S. ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 781~790
DOI : 10.5516/NET.08.2011.062
The HyperGam program was developed for the analysis of complex HPGe
-ray spectra. The previous version of HyperGam was mainly limited to the analysis of
-ray peaks and the manual logging of the result. In this study, it is specifically developed into a tool for the isotopic analysis of spectra. The newly developed features include nuclide identification and activity determination. An algorithm for nuclide identification was developed to identify the peaks in the spectrum by considering the yield, efficiency, energy and peak area for the
-ray lines emitted from the radionuclide. The detailed performance of nuclide identification and activity determination was accessed using the IAEA 2002 set of test spectra. By analyzing the test spectra, the numbers of radionuclides identified truly (true hit), falsely (false hit) or missed (misses) were counted and compared with the results from the IAEA 2002 tests. The determined activities of the radionuclides were also compared for four test spectra of several samples. The result of the performance test is promising in comparison with those of the well-known software packages for
-ray spectrum analysis.
ESTIMATION OF DUCTILE FRACTURE BEHAVIOR INCORPORATING MATERIAL ANISOTROPY
Choi, Shin-Beom ; Lee, Dock-Jin ; Jeong, Jae-Uk ; Chang, Yoon-Suk ; Kim, Min-Chul ; Lee, Bong-Sang ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 791~798
DOI : 10.5516/NET.09.2011.056
Since standardized fracture test specimens cannot be easily extracted from in-service components, several alternative fracture toughness test methods have been proposed to characterize the deformation and fracture resistance of materials. One of the more promising alternatives is the local approach employing the SP(Small Punch) testing technique. However, this process has several limitations such as a lack of anisotropic yield potential and tediousness in the damage parameter calibration process. The present paper investigates estimation of ductile fracture resistance(J-R) curve by FE(Finite Element) analyses using an anisotropic damage model and enhanced calibration procedure. In this context, specific tensile tests to quantify plastic strain ratios were carried out and SP test data were obtained from the previous research. Also, damage parameters constituting the Gurson-Tvergaard-Needleman model in conjunction with Hill's 48 yield criterion were calibrated for a typical nuclear reactor material through a genetic algorithm. Finally, the J-R curve of a standard compact tension specimen was predicted by further detailed FE analyses employing the calibrated damage parameters. It showed a lower fracture resistance of the specimen material than that based on the isotropic yield criterion. Therefore, a more realistic J-R curve of a reactor material can be obtained effectively from the proposed methodology by taking into account a reduced load-carrying capacity due to anisotropy.
IMPROVING REGIONAL OVERPOWER PROTECTION TRIP SET POINT VIA CHANNEL OPTIMIZATION
Kastanya, Doddy ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 799~806
DOI : 10.5516/NET.01.2011.015
In recent years, a new algorithm has been introduced to perform the regional overpower protection (ROP) detector layout optimization for
reactors. This algorithm is called DETPLASA. This algorithm has been shown to successfully come up with a detector layout which meets the target trip set point (TSP) value. Knowing that these ROP detectors are placed in a number of safety channels, one expects that there is an optimal placement of the candidate detectors into these channels. The objective of the present paper is to show that a slight improvement to the TSP value can be realized by optimizing the channelization of these ROP detectors. Depending on the size of the ROP system, based on numerical experiments performed in this study, the range of additional TSP improvement is from 0.16%FP (full power) to 0.56%FP.
UNCERTAINTY AND SENSITIVITY ANALYSIS OF TMI-2 ACCIDENT SCENARIO USING SIMULATION BASED TECHNIQUES
Rao, R. Srinivasa ; Kumar, Abhay ; Gupta, S.K. ; Lele, H.G. ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 807~816
DOI : 10.5516/NET.02.2011.039
The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.
VIBRATION AND STRESS ANALYSIS OF A UGS ASSEMBLY FOR THE APR1400 RVI CVAP
Ko, Do-Young ; Kim, Kyu-Hyung ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 817~824
DOI : 10.5516/NET.09.2011.057
The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.
EDUCATION AND TRAINING IN RADIATION PROTECTION IN KOREA: CURRENT STATUS AND IMPROVEMENTS
Son, Miyeon ; Kim, Hyunkee ; Nam, Youngmi ; Nam, Jongsoo ; Lee, Ki-Bog ;
Nuclear Engineering and Technology, volume 44, issue 7, 2012, Pages 825~830
DOI : 10.5516/NET.10.2011.061
Radiation and its various industrial applications have been growing at approximately 10 percent per year for the past decade in Korea. As a result, the importance of the Education and Training (E&T) in radiation protection is of upmost importance. This paper is intended to investigate the present status of the E&T on radiation protection and safety in Korea and to draw up the improvements of the E&T courses required for building the national radiation safety infrastructure. For these purposes, the E&T data from the six major domestic organizations providing radiation protection training courses were investigated and analyzed. Each of the organizations is offering several kinds of E&T courses based on their own specific functions. These organizations have administrative facilities equipped with the latest technology for E&T in radiation protection. The E&T courses mainly cover the training courses for radiation workers, radiological emergency staff, license applicants, license holders, and regulatory staff. In 2010, a total of 58 E&T courses were carried out across six organizations. The conclusions make a number of observations highlighting challenges such as: establishing a formal feedback mechanism, introducing more practical training sessions, developing training courses tailored to the job categories and target audiences, and designing education and training courses in radiation protection that comply with current obligations as well as future requirements.