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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 43, Issue 6 - Dec 2011
Volume 43, Issue 5 - Oct 2011
Volume 43, Issue 4 - Aug 2011
Volume 43, Issue 3 - Jun 2011
Volume 43, Issue 2 - Apr 2011
Volume 43, Issue 1 - Feb 2011
Selecting the target year
EXTENDED DRY STORAGE OF USED NUCLEAR FUEL: TECHNICAL ISSUES: A USA PERSPECTIVE
Mcconnell, Paul ; Hanson, Brady ; Lee, Moo ; Sorenson, Ken ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 405~412
DOI : 10.5516/NET.2011.43.5.405
Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-tosafety structures, systems, and components (SSCs) to continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSCs. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.
ARISING TECHNICAL ISSUES IN THE DEVELOPMENT OF A TRANSPORTATION AND STORAGE SYSTEM OF SPENT NUCLEAR FUEL IN KOREA
Yoo, Jeong-Hyoun ; Choi, Woo-Seok ; Lee, Sang-Hoon ; Seo, Ki-Seog ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 413~420
DOI : 10.5516/NET.2011.43.5.413
In Korea, although the concept of dry storage system for PWR spent fuels first emerged in the early 1990s, wet storage inside nuclear reactor buildings remains the dominant storage paradigm. Furthermore, as the amount of discharged fuel from nuclear power plants increases, nuclear power plants are confronted with the problem of meeting storage capacity demand. Various measures have been taken to resolve this problem. Dry storage systems along with transportation of spent fuel either on-site or off-site are regarded as the most feasible measure. In order to develop dry storage and transportation system safety analyses, development of design techniques, full scale performance tests, and research on key material degradation should be conducted. This paper deals with two topics, structural analysis methodology to assess cumulative damage to transportation packages and the effects of an aircraft engine crash on a dual purpose cask. These newly emerging issues are selected from among the many technical issues related to the development of transportation and storage systems of spent fuels. In the design process, appropriate analytical methods, procedures, and tools are used in conjunction with a suitably selected test procedure and assumptions such as jet engine simulation for postulated design events and a beyond design basis accident.
OUT-OF-PILE MECHANICAL PERFORMANCE AND MICROSTRUCTURE OF RECRYSTALLIZED ZR-1.5 NB-O-S ALLOYS
Ko, S. ; Lee, J.M. ; Hong, S.I. ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 421~428
DOI : 10.5516/NET.2011.43.5.421
The out-of-pile mechanical performance and microstructure of recrystallized Zr-1.5 Nb-S alloy was investigated. The strength of the recrystallized Zr-1.5Nb-O-S alloys was observed to increase with the addition of sulfur over a wide temperature range, from room temperature up to
. A yield drop and stress serrations due to dynamic strain were observed at room temperature and
. Wavy and curved dislocations and loosely knit tangles were observed after strained to 0.07 at room temperature, suggesting that cross slip is easier. At
, however, dislocations were observed to be straight and aligned along the slip plane, suggesting that cross slip is rather difficult. At
, oxygen atoms are likely to exert a drag force on moving dislocations, intensifying the dynamic strain aging effect. Oxygen atoms segregated at partial dislocations of a screw dislocation with the edge component may hinder the cross slip, resulting in the rather straight dislocations distributed on the major slip planes. Recrystallized Zr-Nb-S alloys exhibited ductile fracture surfaces, supporting the beneficial effect of sulfur in zirconium alloys. Oxidation resistance in air was also found to be improved with the addition of sulfur in Zr-1.5 Nb-O alloys.
INVESTIGATION ON MATERIAL DEGRADATION OF ALLOY 617 IN HIGH TEMPERATURE IMPURE HELIUM COOLANT
Kim, Dong-Jin ; Lee, Gyeong-Geun ; Jeong, Su-Jin ; Kim, Woo-Gon ; Park, Ji-Yeon ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 429~436
DOI : 10.5516/NET.2011.43.5.429
The corrosion of materials exposed to high temperature helium in a very high temperature reactor is caused by interaction with the impurities in the helium. This interaction then induces high temperature mechanical deterioration. By considering the effect of the impurity concentration on material corrosion, a long-term coolant chemistry guideline can be determined for the range of impurity concentration at which the material is stable for a long time. In this work, surface reactions were investigated by analyzing the thermodynamics and the experimental results for Alloy 617 exposed to controlled impure helium at
. Moreover, the surfaces were examined for the Alloy 617 crept in air and in uncontrolled helium, which was explained by possible surface reactions.
PROBABILISTIC APPROACH ON SEISMOGENIC POTENTIAL OF A FAULT
Chang, Chun-Joong ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 437~446
DOI : 10.5516/NET.2011.43.5.437
Siting criteria for nuclear power plants require that faults be characterized as to their potential for generating earthquakes, or that the absence of the potential for these occurrences be demonstrated. Because the definition of active faults in Korea has been applied by the deterministic method, which depends on the numerical age of fault movement, the possibility of inherent uncertainties exists in determining the maximum earthquake from the fault sources for seismic design. In an attempt to overcome these problems this study suggests new criteria and a probabilistic quantitative diagnostic procedure that could estimate whether a fault is capable of generating earthquakes in the near future.
CONTACT FORCE MODEL FOR A BEAM WITH DISCRETELY SPACED GAP SUPPORTS AND ITS APPROXIMATED SOLUTION
Park, Nam-Gyu ; Suh, Jung-Min ; Jeon, Kyeong-Lak ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 447~458
DOI : 10.5516/NET.2011.43.5.447
This paper proposes an approximated contact force model to identify the nonlinear behavior of a fuel rod with gap supports; also, the numerical prediction of interfacial forces in the mechanical contact of fuel rods with gap supports is studied. The Newmark integration method requires the current status of the contact force, but the contact force is not given a priori. Taylor's expansion can be used to predict the unknown contact force; therefore, it should be guaranteed that the first derivative of the contact force is continuous. This work proposes a continuous and differentiable contact force model with the ability to estimate the current state of the contact force. An approximated convex and differentiable potential function for the contact force is described, and a variational formulation is also provided. A numerical example that considers the particularly stiff supports has been studied, and a fuel rod with hardening supports was also examined for a realistic simulation. An approximated proper solution can be obtained using the results, and abrupt changes from the contacting state to non-contacting state, or vice versa, can be relieved. It can also be seen that not only the external force but also the developed contact force affects the response.
PROPERTIES OF LOW-PH CEMENT GROUT AS A SEALING MATERIAL FOR THE GEOLOGICAL DISPOSAL OF RADIOACTIVE WASTE
Kim, Jin-Seop ; Kwon, S. ; Choi, Jong-Won ; Cho, Gye-Chun ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 459~468
DOI : 10.5516/NET.2011.43.5.459
The current solution to the problem of using cementitious material for sealing purposes in a final radioactive waste repository is to develop a low-pH cement grout. In this study, the material properties of a low-pH cement grout based on a recipe used at ONKALO are investigated by considering such factors as pH variation, compressive strength, dynamic modulus, and hydraulic conductivity by using silica fume and micro-cement. From the pH measurements of the hardened cement grout, the required pH (< pH 11) is obtained after 130 days of curing. Although the engineering properties of the low-pH cement grout used in this study are inferior to those of conventional high-pH cement grout, the utilization of silica fume and micro-cement effectively meets the long-term environmental and durability requirements for cement grout in a radioactive waste repository.
ABRASIVE BLASTING TECHNOLOGY FOR DECONTAMINATION OF THE INNER SURFACE OF STEAM GENERATOR TUBES
Kim, Gye-Nam ; Lee, Min-Woo ; Park, Hye-Min ; Choi, Wang-Kyu ; Lee, Kune-Woo ;
Nuclear Engineering and Technology, volume 43, issue 5, 2011, Pages 469~476
DOI : 10.5516/NET.2011.43.5.469
The inner surfaces of bundled inconel tubes from steam generators in South Korean nuclear power plants are contaminated with cobalt and abrasive blasting equipment has been developed to efficiently remove the cobalt. The principal parameters related to the efficient removal using this equipment are the type of abrasive, the distance from the nozzle, and the blasting time. Preliminary tests were performed using oxidized inconel samples which enabled the simulation of cobalt removal from the radioactive inconel samples. The oxygen in the oxidized samples and the cobalt in the radioactive inconel were removed more effectively using the blasting distance, blasting time, and a silicon carbide abrasive. Using the developed abrasive blasting equipment, the optimum decontamination conditions for radioactive inconel samples were blasting for more than 6 minutes using silicon carbides under 5 atmospheric pressures.