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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 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
DEVELOPMENT OF THE ALTERNATE PRESSURIZED THERMAL SHOCK RULE (10 CFR 50.61a) IN THE UNITED STATES
Kirk, Mark ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 277~294
DOI : 10.5516/NET.07.2013.704
In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRC's efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.
CHEMICAL EFFECTS ON PWR SUMP STRAINER BLOCKAGE AFTER A LOSS-OF-COOLANT ACCIDENT: REVIEW ON U.S. RESEARCH EFFORTS
Bahn, Chi Bum ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 295~310
DOI : 10.5516/NET.07.2013.705
Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI)-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET), and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.
IRRADIATION EFFECTS OF HT-9 MARTENSITIC STEEL
Chen, Yiren ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 311~322
DOI : 10.5516/NET.07.2013.706
High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.
CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS
Cheng, Songbai ; Yamano, Hidemasa ; Suzuki, TYohru ; Tobita, Yoshiharu ; Nakamura, Yuya ; Zhang, Bin ; Matsumoto, Tatsuya ; Morita, Koji ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 323~334
DOI : 10.5516/NET.02.2012.068
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
EXPERIMENTAL STUDY OF CRITICAL HEAT FLUX WITH ALUMINA-WATER NANOFLUIDS IN DOWNWARD-FACING CHANNELS FOR IN-VESSEL RETENTION APPLICATIONS
Dewitt, G. ; Mckrell, T. ; Buongiorno, J. ; Hu, L.W. ; Park, R.J. ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 335~346
DOI : 10.5516/NET.02.2012.075
The Critical Heat Flux (CHF) of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR) situation which can occur during core melting sequences in certain advanced Light Water Reactors (LWRs). CHF measurements were conducted in a flow boiling loop featuring a test section designed to be thermal-hydraulically similar to the vessel/insulation gap in the Westinghouse AP1000 plant. The effects of orientation angle, pressure, mass flux, fluid type, boiling time, surface material, and surface state were investigated. Results for water-based nanofluids with alumina nanoparticles (0.001% by volume) on stainless steel surface indicate an average 70% CHF enhancement with a range of 17% to 108% depending on the specific flow conditions expected for IVR. Experiments also indicate that only about thirty minutes of boiling time (which drives nanoparticle deposition) are needed to obtain substantial CHF enhancement with nanofluids.
SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR
Lee, Yeon-Gun ; Park, Il-Woong ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 347~360
DOI : 10.5516/NET.02.2013.024
This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.
VISUALIZATION OF INTERNAL DEFECTS IN PLATE-TYPE NUCLEAR FUEL BY USING NONCONTACT OPTICAL INTERFEROMETRY
Park, Seung-Kyu ; Park, Nak-Gyu ; Baik, Sung-Hoon ; Kang, Young-June ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 361~366
DOI : 10.5516/NET.05.2012.065
An imaging technique to visualize the internal defects in a plate-type nuclear fuel specimen was developed by using an active optical interferometer for a nondestructive quality inspection. A periodic thermal wave having a sinusoidal intensity pattern induced a periodical strain variation for the specimen. The varying strain image was acquired using an optical laser interferometer. The strain distribution over the internal defects will be distorted in an acquired strain image because a part of the thermal wave will be reflected from these defects during propagation. In this paper, internal defects were efficiently visualized by sequentially accumulating the extracted defect components. The experimental results confirmed that the developed visualization system can be a valuable tool to detect the internal defects in plate-type nuclear fuel.
A SCENARIO STUDY ON MIXING STRATEGIES OF FAST REACTOR WITH LOW AND HIGH CONVERSION RATIOS
Jeong, Chang Joon ; Jo, Chang Keun ; Noh, Jae Man ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 367~376
DOI : 10.5516/NET.06.2012.031
This study investigated mixing scenarios of the low and high conversion ratios (CRs) of fast reactors (FRs). The fuel cycle was modeled so as to minimize the spent fuel (SF) or transuranics (TRU) inventories. The scenarios were modeled for a single low CR of 0.61 and a high CR of 1.0. The study also investigated the mixing scenario of low-high CR and/or high-low CR. The SF and TRU inventories, associated with different scenarios, were compared to those of the light water reactor (LWR) once-through (OT) case. Also, the important isotope concentration and long-term heat (LTH) load were calculated and compared to those of the OT cycle. As a result, it is known that the deployment of FRs of low CR burns more TRU and results in a reduction of the out-of-pile TRU inventory and LTH with low deployment capacity. This study shows that the mixing strategy of FRs of low and high CR can reduce the SF and TRU inventories with lower deployment capacity as compared with a single deployment of FRs of high CR.
A STUDY ON THE INITIAL CHARACTERISTICS OF DOMESTIC SPENT NUCLEAR FUELS FOR LONG TERM DRY STORAGE
Kim, Juseong ; Yoon, Hakkyu ; Kook, Donghak ; Kim, Yongsoo ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 377~384
DOI : 10.5516/NET.06.2012.082
During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-
as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from
, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100
, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.
RADIATION-INDUCED DISLOCATION AND GROWTH BEHAVIOR OF ZIRCONIUM AND ZIRCONIUM ALLOYS - A REVIEW
Choi, Sang Il ; Kim, Ji Hyun ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 385~392
DOI : 10.5516/NET.07.2013.035
Zirconium and zirconium alloys are widely used as nuclear reactor core materials such as fuel cladding and guide tubes because they have excellent corrosion- and radiation-resistant properties. In the reactor core, zirconium alloys are subjected to high-energy neutron fluence, causing radiation-induced dislocation and growth. To discern a possible correlation between radiation-induced dislocation and growth, characteristics of dislocation and growth in zirconium and its alloys are examined. The radiation-induced dislocation including
dislocation loops is reviewed in various temperature and fluence ranges, and their growth behavior is examined in the same way. To further a fundamental understanding, radiation-induced growth prediction models are briefly reviewed. This research will assist in the design of zirconium based components as well as the safety analysis of various reactor conditions, in both research and commercial reactors.
AN EXPERIMENTAL INVESTIGATION ON MINIMUM COMPRESSIVE STRENGTH OF EARLY AGE CONCRETE TO PREVENT FROST DAMAGE FOR NUCLEAR POWER PLANT STRUCTURES IN COLD CLIMATES
Koh, Kyung-Taek ; Park, Chun-Jin ; Ryu, Gum-Sung ; Park, Jung-Jun ; Kim, Do-Gyeum ; Lee, Jang-Hwa ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 393~400
DOI : 10.5516/NET.09.2012.046
Concrete undergoing early frost damage in cold weather will experience significant loss of not only strength, but also of permeability and durability. Accordingly, concrete codes like ACI-306R prescribe a minimum compressive strength and duration of curing to prevent frost damage at an early age and secure the quality of concrete. Such minimum compressive strength and duration of curing are mostly defined based on the strength development of concrete. However, concrete subjected to frost damage at early age may not show a consistent relationship between its strength and durability. Especially, since durability of concrete is of utmost importance in nuclear power plant structures, this relationship should be imperatively clarified. Therefore, this study verifies the feasibility of the minimum compressive strength specified in the codes like ACI-306R by evaluating the strength development and the durability preventing the frost damage of early age concrete for nuclear power plant. The results indicate that the value of 5 MPa specified by the concrete standards like ACI-306R as the minimum compressive strength to prevent the early frost damage is reasonable in terms of the strength development, but seems to be inappropriate in the viewpoint of the resistance to chloride ion penetration and freeze-thaw. Consequently, it is recommended to propose a minimum compressive strength preventing early frost damage in terms of not only the strength development, but also in terms of the durability to secure the quality of concrete for nuclear power plants in cold climates.
A SMALL MODULAR REACTOR DESIGN FOR MULTIPLE ENERGY APPLICATIONS: HTR50S
Yan, X. ; Tachibana, Y. ; Ohashi, H. ; Sato, H. ; Tazawa, Y. ; Kunitomi, K. ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 401~414
DOI : 10.5516/NET.10.2012.070
HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's
, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to
for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to
for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.
INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY FOR THE DETERMINATION OF
Np IN SPENT NUCLEAR FUEL SAMPLES BY ISOTOPE DILUTION METHOD USING
Np AS A SPIKE
Joe, Kihsoo ; Han, Sun-Ho ; Song, Byung-Chul ; Lee, Chang-Heon ; Ha, Yeong-Keong ; Song, Kyuseok ;
Nuclear Engineering and Technology, volume 45, issue 3, 2013, Pages 415~420
DOI : 10.5516/NET.06.2012.056
A determination method for
in spent nuclear fuel samples was developed using an isotope dilution method with
as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the
instead of the previously used alpha spectrometry.
were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of
in synthetic samples was
% (1S, n=4). The
contents in the spent fuel samples were 0.15, 0.25, and
and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively.