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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 35, Issue 6 - Dec 2003
Volume 35, Issue 5 - Oct 2003
Volume 35, Issue 4 - Aug 2003
Volume 35, Issue 3 - Jun 2003
Volume 35, Issue 2 - Apr 2003
Volume 35, Issue 1 - Feb 2003
Selecting the target year
Non-Destructive Detection of Hydride Blister in PHWR Pressure Tube Using an Ultrasonic Velocity Ratio Method
Cheong Yong-Moo ; Lee Dong-Hoon ; Kim Sang-Jae ; Kim Young-Suk ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 369~377
Since Zr-2.5Nb pressure tubes have a high risk for the formation of blisters during their operation in pressurized heavy water reactors, there has been a strong incentive to develop a method for the non-destructive detection of blisters grown on the tube surfaces. However, because there is little mismatch in acoustic impedance between the hydride blisters and zirconium matrix, it is not easy to distinguish the boundary between the blister and zirconium matrix with conventional ultrasonic methods. This study has focused on the development of a special ultrasonic method, so called ultrasonic velocity ratio method for a reliable detection of blisters formed on Zr-2.5Nb pressure tubes. Hydride blisters were grown on the outer surface of the Zr-2.5Nb pressure tube using a cold finger attached to a steady state thermal diffusion equipment. To maximize a difference in the ultrasonic velocity in hydride blisters and the zirconium matrix, the ultrasonic velocity ratio of longitudinal wave to shear wave,
, has been determined based on the flight time of the longitudinal echo and reflected shear echo from the outer surface of the tubes. The feasibility of the ultrasonic velocity ratio method is confirmed by comparing the contour plots reproduced by this method with those of the blisters grown on the Zr-2.5Nb pressure tubes.
Hydriding Failure Analysis Based on PIE Data
Kim Yong-Soo ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 378~386
Recently failures of nuclear fuel rods in Korean nuclear power plants were reported and their failure causes have been investigated by using PIE techniques. Destructive and physico-chemical examinations reveal that the clad hydriding phenomena had caused the rod failures primarily and secondarily in each case. In this study, the basic mechanisms of the primary and the secondary hydriding failures are reviewed, PIE data such as cladding inner and outer surface oxide thickness and the restructuring of the fuel pellets are analyzed, and they are compared with the predicted behaviors by a fuel performance code. In addition, post-defected fuel behaviors are reviewed and qualitatively analyzed. The results strongly support that the hydriding processes, primary and secondary, played critical roles in the respective fuel rods failures and the secondary hydriding failure can take place even in the fuel rod with low linear heat generation rate.
Statistical Approach for Derivation of Quantitative Acceptance Criteria for Radioactive Wastes to Near Surface Disposal Facility
Park Jin Beak ; Park Joo Wan ; Lee Eun Yong ; Kim Chang Lak ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 387~398
For reference human intrusion scenarios constructed in previous study, a probabilistic safety assessment to derive the radionuclide concentration limits for the low- and intermediate- level radioactive waste disposal facility is conducted. Statistical approach by the Latin Hypercube Sampling method is introduced and new assumptions about the disposal facility system are examined and discussed. In our previous study of deterministic approach, the post construction scenarios appeared as most limiting scenario to derive the radionuclide concentration limits. Whereas, in this statistical approach, the post drilling and the post construction scenarios are mutually competing for the scenario selection according to which radionuclides are more important in safety assessment context. Introduction of new assumption shows that the post drilling scenario can play an important role as the limiting scenario instead of the post-construction scenario. When we compare the concentration limits between the previous and this study, concentrations of radionuclides such as Nb-94, Cs-137 and alpha-emitting radionuclides show elevated values than the case of the previous study. Remaining radionuclides such as Sr-90, Tc-99 I-129, Ni-59 and Ni-63 show lower values than the case of the previous study.
A Model Predictive Controller for Nuclear Reactor Power
Na Man Gyun ; Shin Sun Ho ; Kim Whee Cheol ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 399~411
A model predictive control method is applied to design an automatic controller for thermal power control in a reactor core. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the second optimal control input is not implemented and the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize the difference between the output and the desired output and the variation of the control rod position. The nonlinear PWR plant model (a nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) is used to verify the proposed controller of reactor power. And a controller design model used for designing the model predictive controller is obtained by applying a parameter estimation algorithm at an initial stage. From results of numerical simulation to check the controllability of the proposed controller at the
ramp increase or decrease of a desired load and its
step increase or decrease which are design requirements, the performances of this controller are proved to be excellent.
The Operators' Non-compliance Behavior to Conduct Emergency Operating Procedures - Comparing with the Complexity of the Procedural Steps
Park Jinkyun ; Jung Wondea ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 412~425
According to the results of related studies, one of the typical factors related to procedure related human errors is the complexity of procedures. This means that comparing the change of the operators' behavior with respect to the complexity of procedures may be meaningful in clarifying the reasons for the operators' non-compliance behavior. In this study, to obtain data related to the operators' non-compliance behavior, emergency training records were collected using a full scope simulator. And three types of the operators' behavior (such as strict adherence, skipping redundant actions and modifying action sequences) observed from the collected emergency training records were compared with the complexity of the procedural steps. As the results, two remarkable relationships are obtained. They are: 1) the operators seem to frequently adopt non-compliance behavior to conduct the procedural steps that have an intermediate procedural complexity, 2) the operators seems to accommodate their non-compliance behavior to the complexity of the procedural steps. Therefore, it is expected that these relationships can be used as meaningful clues not only to scrutinize the reason for non-compliance behavior but also to suggest appropriate remedies for the reduction of non-compliance behavior that can result in procedure related human error.
Assessments of RELAP5/MOD3.2 and RELAP5/CANDU in a Reactor Inlet Header Break Experiment B9401 of RD-14M
Cho Yong Jin ; Jeun Gyoo Dong ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 426~441
A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.
Fretting-Wear Characteristics of Steam Generator Tubes by Foreign Object
Jo Jong Chull ; Jhung Myung Jo ; Kim Woong Sik ; Choi Young Hwan ; Kim Hho Jung ; Kim Tae Hyung ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 442~453
This study investigates the safety assessment of the potential for fretting-wear damages on steam generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element modelings of U-tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. Also, discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube.
Fault Diagnosis for Agitator Driving System in a High Temperature Reduction Reactor
Park Gee Young ; Hong Dong Hee ; Jung Jae Hoo ; Kim Young Hwan ; Jin Jae Hyun ; Yoon Ji Sup ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 454~470
In this paper, a preliminary study for development of a fault diagnosis is presented for monitoring and diagnosing faults in the agitator driving system of a high temperature reduction reactor. In order to identify a fault occurrence and classify the fault cause, vibration signals measured by accelerometers on the outer shroud of the agitator driving system are firstly decomposed by wavelet transform (WT) and the features corresponding to each fault type are extracted. For the diagnosis, the fuzzy ARTMAP is employed and thereby, based on the features extracted from the WT, the robust fault classifier can be implemented with a very short training time - a single training epoch and a single learning iteration is sufficient for training the fault classifier. The test results demonstrate satisfactory classification for the faults pre-categorized from considerations of possible occurrence during experiments on a small-scale reduction reactor.
Microstructural Characteristics of the Fuel Cladding Tubes Irradiated in Kori Unit 1
Kim H.G. ; Baek J.H. ; Lee M.H. ; Chun Y.B. ; Jeong Y.H. ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 471~481
In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from
and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly
for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both
-type dislocation loops and
-type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.
Low & Intermediate Level Radioactive Waste Vitrification Using Plasma Arc Melting Technology
Min Byeong-Yeon ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 482~496
effectiveness of the PAM graphite-electrode technology for the treatment of many types of low-level radioactive waste including : combustible material, solidified resins in cement, inorganic materials, steel, glass, and solidified boric acid cement. The objectives of PAM-200 evaluation were to verify that 1) the facility meets air emission regulations, 2) the facility can be safely operated when processing hazardous and radioactive materials and 3) satisfactory final waste forms can be produced. Results, derived from KAERI's(Korea Atomic Energy Research Institute) analyses for samples of vitrified product, scrubbing solution and offgas collected during test period, show that PAM-200 can treat radioactive wastes as well as hazardous wastes with toxic constituents and radionuclides contained in the offgas exiting from the stack to the environment controlled to be far lower than the limit regulated by air conservation law and atomic law.
Effect of Neutron Energy Spectra on the Formation of the Displacement Cascade in
Kwon Junhyun ; Seo Chul Gyo ; Kwon Sang Chul ; Hong Jun-Hwa ;
Nuclear Engineering and Technology, volume 35, issue 5, 2003, Pages 497~505
This paper describes a computational approach to the quantification of primary damage under irradiation and demonstrates the effect of neutron energy spectra on the formation of the displacement cascade. The development of displacement cascades in
has been simulated using the MOLDY code - a molecular dynamics code for simulating radiation damage. The primary knock-on atom energy, key input to the MOLDY code, was determined from the SPECTER code calculation on two neutron spectra. The two neutron spectra include; (i) neutron spectrum in the instrumented irradiation capsule of the high-flux advanced neutron application reactor (HANARO), and (ii) neutron spectrum at the inner surface of the reactor pressure vessel steel for the Younggwang nuclear power plant No.5 (YG 5). Minor differences in the normalized neutron spectra between the two spectra produce similar values of PKA energy, which are 4.7 keV for HANARO and 5.3 keV for YG 5. This similarity implies that primary damage to the components of the commercial nuclear reactors should be well simulated by irradiation in the HANARO. Moreover, the application of the MD calculations corroborates this statement by comparing cascades simulation results.