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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 40, Issue 7 - Dec 2008
Volume 40, Issue 6 - Oct 2008
Volume 40, Issue 5 - Aug 2008
Volume 40, Issue 4 - Jun 2008
Volume 40, Issue 3 - Apr 2008
Volume 40, Issue 2 - Mar 2008
Volume 40, Issue 1 - Feb 2008
Selecting the target year
SOME POWER UPRATE ISSUES IN NUCLEAR POWER PLANTS
Tipping, Philip ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 251~254
DOI : 10.5516/NET.2008.40.4.251
Issues and themes concerned with nuclear power plant uprating are examined. Attention is brought to the fact that many candidate nuclear power plants for uprating have anyway been operated below their rated power for a significant part of their operating life. The key issues remain safety and reliability in operation at all times, irrespective of the nuclear power plant's chronological or design age or power rating. The effects of power uprates are discussed in terms of material aspects and expected demands on the systems, structures and components. The impact on operation and maintenance methods is indicated in terms of changes to the ageing surveillance programmes. Attention is brought to the necessity checking or revising operator actions after power up-rating has been implemented.
POWER UPRATES IN NUCLEAR POWER PLANTS: INTERNATIONAL EXPERIENCES AND APPROACHES FOR IMPLEMENTATION
Kang, Ki-Sig ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 255~268
DOI : 10.5516/NET.2008.40.4.255
The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.
EXTENSION OF OPERATIONAL LIFE-TIME OF WWER-440/213 TYPE UNITS AT PAKS NUCLEAR POWER PLANT
Katona, Tamas Janos ; Ratkai, Sandor ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 269~276
DOI : 10.5516/NET.2008.40.4.269
Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.
TECHNICAL EVALUATION OF THE CONTINUED OPERATION OF NPP
Kim, Tae-Ryong ; Jin, Tae-Eun ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 277~284
DOI : 10.5516/NET.2008.40.4.277
Recently, the long-term operation of a nuclear power plant beyond its licensed term has become a worldwide trend as long as the safety of the plant is maintained in the extended period. Kori Unit 1, the oldest PWR in Korea, is the foremost example of this type of long-term operation in Korea. Comprehensive technical evaluation of the long-term operation of this plant was completed to confirm the overall safety of the plant. The technical evaluation included a review of PSR results, an assessment on aging management programs and time limited aging analyses, and a statement of radiological impact on the environment. Based on all of the results of the technical evaluation activities, Kori Unit 1 was approved to operate for an additional 10 years beyond its original design life of 30 years.
A PERFORMANCE ASSESSMENT OF A BASE ISOLATION SYSTEM FOR AN EMERGENCY DIESEL GENERATOR IN A NUCLEAR POWER PLANT
Choun, Young-Sun ; Kim, Min-Kyu ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 285~298
DOI : 10.5516/NET.2008.40.4.285
This study evaluates the performance of a coil spring-viscous damper system for the vibration and seismic isolation of an Emergency Diesel Generator (EDG) by measuring its operational vibration and seismic responses. The vibration performance of a coil spring-viscous damper system was evaluated by the vibration measurements for an identical EDG set with different base systems - one with an anchor bolt system and the other with a coil spring-viscous damper system. The seismic performance of the coil spring-viscous damper system was evaluated by seismic tests with a scaled model of a base-isolated EDG on a shaking table. The effects of EDG base isolation on the fragility curve and core damage frequency in a nuclear power plant were also investigated through a case study.
MEASUREMENT OF THE D-D NEUTRON GENERATION RATE BY PROTON COUNTING
Kim, In-Jung ; Jung, Nam-Suk ; Choi, Hee-Dong ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 299~304
DOI : 10.5516/NET.2008.40.4.299
A detection system was set up to measure the neutron generation rate of a recently developed D-D neutron generator. The system is composed of a Si detector, He-3 detector, and electronics for pulse height analysis. The neutron generation rate was measured by counting protons using the Si detector, and the data was crosschecked by counting neutrons with the He-3 detector. The efficiencies of the Si and He-3 detectors were calibrated independently by using a standard alpha particle source
and a bare isotopic neutron source
, respectively. The effect of the cross-sectional difference between the D(d,p)T and
reactions was evaluated for the case of a thick target. The neutron generation rate was theoretically corrected for the anisotropic emission of protons and neutrons in the D-D reactions. The attenuations of neutron on the path to the He-3 detector by the target assembly and vacuum flange of the neutron generator were considered by the Monte Carlo method using the MCNP 4C2 code. As a result, the neutron generation rate based on the Si detector measurement was determined with a relative uncertainty of
, and the two rates measured by both detectors corroborated within 20%.
MICROSTRUCTURAL OBSERVATION AND TENSILE ISOTROPY OF AN AUSTENITIC ODS STEEL
Kim, Tae-Kyu ; Bae, Chang-Soo ; Kim, Do-Hyang ; Jang, Jin-Sung ; Kim, Sung-Ho ; Lee, Chan-Bock ; Hahn, Do-Hee ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 305~310
DOI : 10.5516/NET.2008.40.4.305
Based on a composition of 99.4 wt% AISI 316L stainless steel, 0.3wt% Ti and 0.3 wt%
, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling. Fine oxide particles were observed in the matrix, and their chemical formulations were determined to be
and TiO. Heat treatment of the cold-rolled sample at
induced an isotropic tensile behavior at room temperature and at
. This result would be mainly attributed to the equiaxed grains that form as a result of the heat treatment for recrystallization.
SHIELDED LASER ABLATION ICP-MS SYSTEM FOR THE CHARACTERIZATION OF HIGH BURNUP FUEL
Ha, Yeong-Keong ; Han, Sun-Ho ; Kim, Hyun-Gyum ; Kim, Won-Ho ; Jee, Kwang-Yong ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 311~318
DOI : 10.5516/NET.2008.40.4.311
In modem power reactors, nuclear fuels have recently reached 55,000 MWd/MtU from the initial average burnup of 35,000 MWd/MtU to reduce the fuel cycle cost and waste volume. At such high burnups, a fuel pellet produces fission products proportional to the burnup and creates a typical high burnup structure around the periphery region of the pellet, producing the so called 'rim effect'. This rim region of a highly burnt fuel is known to be ca.
in width and is known to affect the fuel integrity. To characterize the local burnup in the rim region, solid sampling in the micro meter region by laser ablation is needed so that the distribution of isotopes can be determined by ICP-MS. For this procedure, special radiation shielding is required for personnel safety. In this study, we installed a radiation shielded laser ablation ICP-MS system, and a performance test of the developed system was conducted to evaluate the safe operation of instruments.
SHIELDING PERFORMANCE OF A NEWLY DESIGNED TRANSPORT CASK IN THE ADVANCED CONDITIONING SPENT FUEL PYROPROCESS FACILITIY
Park, Chang-Je ; Jeong, Chang-Joon ; Min, Deok-Ki ; Kang, Hee-Young ; Choi, Woo-Seok ; Lee, Joo-Chan ; Bang, Gyeoung-Sik ; Seo, Ki-Seog ;
Nuclear Engineering and Technology, volume 40, issue 4, 2008, Pages 319~326
DOI : 10.5516/NET.2008.40.4.319
To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).