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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 45, Issue 7 - Dec 2013
Volume 45, Issue 6 - Nov 2013
Volume 45, Issue 5 - Oct 2013
Volume 45, Issue 4 - Aug 2013
Volume 45, Issue 3 - Jun 2013
Volume 45, Issue 2 - Apr 2013
Volume 45, Issue 1 - Feb 2013
Selecting the target year
MAKING THE CASE FOR SAFE STORAGE OF USED NUCLEAR FUEL FOR EXTENDED PERIODS OF TIME: COMBINING NEAR-TERM EXPERIMENTS AND ANALYSES WITH LONGER-TERM CONFIRMATORY DEMONSTRATIONS
Sorenson, Ken B. ; Hanson, Brady ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 421~426
DOI : 10.5516/NET.06.2013.707
The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel. This paper discusses a technical approach that defines a process that develops the technical basis for demonstrating the safety of used fuel over extended periods of time.
CONSIDERATIONS REGARDING ROK SPENT NUCLEAR FUEL MANAGEMENT OPTIONS
Braun, Chaim ; Forrest, Robert ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 427~438
DOI : 10.5516/NET.06.2013.708
In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, will not suffice to handle all the spent fuel discharged annually. Hence centralized dry cask storage will be required even if the pyroprocessing option is successfully developed by 2030. Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR). We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-fromreactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.
TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR
Lee, Yeon-Gun ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 439~458
DOI : 10.5516/NET.02.2013.025
REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.
ESTABLISHMENT OF A SEVERE ACCIDENT MITIGATION STRATEGY FOR AN SBO AT WOLSONG UNIT 1 NUCLEAR POWER PLANT
Kim, Sungmin ; Kim, Dongha ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 459~468
DOI : 10.5516/NET.03.2012.057
During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.
THERMAL AND STRUCTURAL ANALYSIS OF CALANDRIA VESSEL OF A PHWR DURING A SEVERE ACCIDENT
Kulkarni, P.P. ; Prasad, S.V. ; Nayak, A.K. ; Vijayan, P.K. ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 469~476
DOI : 10.5516/NET.03.2012.052
In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.
A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA
Yoo, Junbeom ; Lee, Jong-Hoon ; Lee, Jang-Soo ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 477~488
DOI : 10.5516/NET.04.2012.078
The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.
A BEHAVIOR-PRESERVING TRANSLATION FROM FBD DESIGN TO C IMPLEMENTATION FOR REACTOR PROTECTION SYSTEM SOFTWARE
Yoo, Junbeom ; Kim, Eui-Sub ; Lee, Jang-Soo ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 489~504
DOI : 10.5516/NET.04.2012.085
Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation.
INFLUENCE OF ALLOY COMPOSITION ON WORK HARDENING BEHAVIOR OF ZIRCONIUM-BASED ALLOYS
Kim, Hyun-Gil ; Kim, Il-Hyun ; Park, Jeong-Yong ; Koo, Yang-Hyun ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 505~512
DOI : 10.5516/NET.07.2012.055
Three types of zirconium base alloy were evaluated to study how their work hardening behavior is affected by alloy composition. Repeated-tensile tests (5% elongation at each test) were performed at room temperature at a strain rate of
for the alloys, which were initially controlled for their microstructure and texture. After considering the yield strength and work hardening exponent (n) variations, it was found that the work hardening behavior of the zirconium base alloys was affected more by the Nb content than the Sn content. The facture mode during the repeated tensile test was followed by the slip deformation of the zirconium structure from the texture and microstructural analysis.
A REVIEW ON THE ODSCC OF STEAM GENERATOR TUBES IN KOREAN NPPS
Chung, Hansub ; Kim, Hong-Deok ; Oh, Seungjin ; Boo, Myung Hwan ; Na, Kyung-Hwan ; Yun, Eunsup ; Kang, Yong-Seok ; Kim, Wang-Bae ; Lee, Jae Gon ; Kim, Dong-Jin ; Kim, Hong Pyo ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 513~522
DOI : 10.5516/NET.07.2012.090
The ODSCC detected in the TSP position of Ulchin 3&4 SGs are typical ODSCC of Alloy 600MA tubes. The causative chemical environment is formed by concentration of impurities inside the occluded region formed by the tube surface, egg crate strips, and sludge deposit there. Most cracks are detected at or near the line contacts between the tube surface and the egg crate strips. The region of dense crack population, as defined as between
TSPs, and near the center of hot leg hemisphere plane, coincided well with the region of preferential sludge deposition as defined by thermal hydraulics calculation using SGAP computer code. The cracks developed homogeneously in a wide range of SGs, so that the number of cracks detected each outage increased very rapidly since the first detection in the
refueling outage. The root cause assessment focused on investigation of the difference in microstructure and manufacturing residual stress in order to reveal the cause of different susceptibilities to ODSCC among identical six units. The manufacturing residual stress as measured by XRD on OD surface and by split tube method indicated that the high residual stress of Alloy 600MA tube played a critical role in developing ODSCC. The level of residual stress showed substantial variations among the six units depending on details of straightening and OD grinding processes. Youngwang 3&4 tubes are less susceptible to ODSCC than U3 and U4 tubes because semi-continuous coarse chromium carbides are formed along the grain boundary of Y3&4 tubes, while there are finer less continuous chromium carbides in U3 and U4. The different carbide morphology is caused by the difference in cooling rate after mill anneal. There is a possibility that high chromium content in the Y3&4 tubes, still within the allowable range of Alloy 600, has made some contribution to the improved resistance to ODSCC. It is anticipated that ODSCC in Y5&6 SGs will be retarded more considerably than U3 SGs since the manufacturing residual stress in Y5&6 tubes is substantially lower than in U3 tubes, while the microstructure is similar with each other.
DESIGN AND FABRICATION OF THE BEAM POSITION MONITOR FOR THE PEFP LINAC
Kwon, Hyeok-Jung ; Kim, Han-Sung ; Seol, Kyung-Tae ; Ryu, Jin-Yeong ; Jang, Ji-Ho ; Cho, Yong-Sub ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 523~528
DOI : 10.5516/NET.08.2012.059
The beam position monitor (BPM) is an essential component for the PEFP 100-MeV linac's commissioning. A prototype stripline-type linac BPM was designedfor this purpose. The electrode aperture is 20 mm in diameter, and the electrode is 25 mm long, so it can be installed between Drift Tube Linac (DTL)101 and DTL102, which is the shortest distance. One end of the electrode is connected to the Sub Miniature Type A (SMA) feed through for signal measurement, and the other end is terminated as a short. The signal amplitude of the fundamental component was calculated and compared with that of the second harmonic component. The designed BPM was fabricated and a low-power RF test was conducted. In this paper, the design, fabrication and low power test of the BPM for the PEFP linac are presented.
SCATTERING CORRECTION FOR IMAGE RECONSTRUCTION IN FLASH RADIOGRAPHY
Cao, Liangzhi ; Wang, Mengqi ; Wu, Hongchun ; Liu, Zhouyu ; Cheng, Yuxiong ; Zhang, Hongbo ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 529~538
DOI : 10.5516/NET.08.2012.074
Scattered photons cause blurring and distortions in flash radiography, reducing the accuracy of image reconstruction significantly. The effect of the scattered photons is taken into account and an iterative deduction of the scattered photons is proposed to amend the scattering effect for image restoration. In order to deduct the scattering contribution, the flux of scattered photons is estimated as the sum of two components. The single scattered component is calculated accurately together with the uncollided flux along the characteristic ray, while the multiple scattered component is evaluated using correction coefficients pre-obtained from Monte Carlo simulations.The arbitrary geometry pretreatment and ray tracing are carried out based on the customization of AutoCAD. With the above model, an Iterative Procedure for image restORation code, IPOR, is developed. Numerical results demonstrate that the IPOR code is much more accurate than the direct reconstruction solution without scattering correction and it has a very high computational efficiency.
FACTORS OF GROUNDWATER FLUCTUATION IN SHIN KORI NUCLEAR POWER PLANTS IN KOREA
Hyun, Seung Gyu ; Woo, Nam C. ; Kim, Kue-Young ; Lee, Hyun-A ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 539~552
DOI : 10.5516/NET.09.2012.072
To establish an aging management plan considering seawater influx and changes in groundwater within nuclear power plant sites, the characteristics of groundwater flow must be understood. This study investigated the characteristics of groundwater flow within the site and analyzed groundwater level recorded by monitoring wells to evaluate groundwater flow characteristics and elements that affected these characteristics for supplying the information to conduct the appropriate aging management for ensuring the safety of the safety-related structures in Shin Kori Unit 1 and 2. The increase in groundwater level during the wet season results from high sea-level conditions and the large amount of precipitation. As a result of the analysis of groundwater distribution and change characteristics, the site could be divided into a rainfall-affected area and a tide-affected area. First, the rainfall-affected area can further be divided into areas that are affected simultaneously by excavation, backfill, and a permanent dewatering system. Secondly, areas that are not affected by excavation, or the dewatering system, or by structure arrangement and excavation. Analysis of the spectrum for wells affected by tides resulted in confirmation of the M2 component (12.421 hr) and S2 component (12.000 hr) of the semidiurnal tides, and the O1 component (25.819 hr) of the diurnal tides. In the cross-correlation results regarding tides and groundwater levels, the lag time occurred diversely within 1-3 hours by the effect of the well location from sea, the distribution of the backfill material with depth, and the concrete structure.
LOCAL COLLISION SIMULATION OF AN SC WALL USING ENERGY ABSORBING STEEL
Chung, Chul-Hun ; Choi, Hyun ; Park, Jaegyun ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 553~564
DOI : 10.5516/NET.09.2013.007
This study evaluates the local damage of a turbine in an auxiliary building of a nuclear power plant due to an external impact by using the LS-DYNA finite element program. The wall of the auxiliary building is SC structure and the material of the SC wall plate is high manganese steel, which has superior ductility and energy absorbance compared to the ordinary steel used for other SC wall plates. The effects of the material of the wall, collision speed, and angle on the magnitude of the local damage were evaluated by local collision analysis. The analysis revealed that the SC wall made of manganese steel had significantly less damage than the SC wall made of ordinary steel. In conclusion, an SC wall made of manganese steel can have higher effective resistance than an SC wall made of ordinary steel against the local collision of an airplane engine or against a turbine impact.
FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING
Kim, Weon-Ju ; Kim, Daejong ; Park, Ji Yeon ;
Nuclear Engineering and Technology, volume 45, issue 4, 2013, Pages 565~572
DOI : 10.5516/NET.07.2012.084
The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade
composites are briefly reviewed. A CVI-processed
composite with a PyC or
interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.