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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
THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS
Kastanya, D. ; Boyle, S. ; Hopwood, J. ; Park, Joo Hwan ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 573~580
DOI : 10.5516/NET.03.2013.709
The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The
reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.
SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU
6 NUCLEAR REACTORS
Hartmann, Wolfgang ; Jung, Jong Yeob ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 581~588
DOI : 10.5516/NET.03.2013.710
This paper deals with the Safety Analysis for
6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.
BACKUP AND ULTIMATE HEAT SINKS IN CANDU REACTORS FOR PROLONGED SBO ACCIDENTS
Nitheanandan, T. ; Brown, M.J. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 589~596
DOI : 10.5516/NET.03.2013.711
In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.
OVERVIEW OF CONTAINMENT FILTERED VENT UNDER SEVERE ACCIDENT CONDITIONS AT WOLSONG NPP UNIT 1
Song, Y.M. ; Jeong, H.S. ; Park, S.Y. ; Kim, D.H. ; Song, J.H. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 597~604
DOI : 10.5516/NET.03.2013.712
Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.
MANAGING A PROLONGED STATION BLACKOUT CONDITION IN AHWR BY PASSIVE MEANS
Kumar, Mukesh ; Nayak, A.K. ; Jain, V ; Vijayan, P.K. ; Vaze, K.K. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 605~612
DOI : 10.5516/NET.02.2012.086
Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000
, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.
IDENTIFICATION OF TWO-DIMENSIONAL VOID PROFILE IN A LARGE SLAB GEOMETRY USING AN IMPEDANCE MEASUREMENT METHOD
Euh, D.J. ; Kim, S. ; Kim, B.D. ; Park, W.M. ; Kim, K.D. ; Bae, J.H. ; Lee, J.Y. ; Yun, B.J. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 613~624
DOI : 10.5516/NET.02.2013.023
Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of
. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.
ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR
Jain, Vikas ; Nayak, A.K. ; Dhiman, M. ; Kulkarni, P.P. ; Vijayan, P.K. ; Vaze, K.K. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 625~636
DOI : 10.5516/NET.03.2013.013
Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.
AN ANALYSIS OF TECHNICAL SECURITY CONTROL REQUIREMENTS FOR DIGITAL I＆C SYSTEMS IN NUCLEAR POWER PLANTS
Song, Jae-Gu ; Lee, Jung-Woon ; Park, Gee-Yong ; Kwon, Kee-Choon ; Lee, Dong-Young ; Lee, Cheol-Kwon ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 637~652
DOI : 10.5516/NET.04.2012.091
Instrumentation and control systems in nuclear power plants have been digitalized for the purpose of maintenance and precise operation. This digitalization, however, brings out issues related to cyber security. In the most recent past, international standard organizations, regulatory institutes, and research institutes have performed a number of studies addressing these systems cyber security.. In order to provide information helpful to the system designers in their application of cyber security for the systems, this paper presents methods and considerations to define attack vectors in a target system, to review and select the requirements in the Regulatory Guide 5.71, and to integrate the results to identify applicable technical security control requirements. In this study, attack vectors are analyzed through the vulnerability analyses and penetration tests with a simplified safety system, and the elements of critical digital assets acting as attack vectors are identified. Among the security control requirements listed in Appendices B and C to Regulatory Guide 5.71, those that should be implemented into the systems are selected and classified in groups of technical security control requirements using the results of the attack vector analysis. For the attack vector elements of critical digital assets, all the technical security control requirements are evaluated to determine whether they are applicable and effective, and considerations in this evaluation are also discussed. The technical security control requirements in three important categories of access control, monitoring and logging, and encryption are derived and grouped according to the elements of attack vectors as results for the sample safety system.
METHOD FOR THE ANALYSIS OF TEMPORAL CHANGE OF PHYSICAL STRUCTURE IN THE INSTRUMENTATION AND CONTROL LIFE-CYCLE
Goring, Markus ; Fay, Alexander ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 653~664
DOI : 10.5516/NET.04.2013.010
The design of computer-based instrumentation and control (I&C) systems is determined by the allocation of I&C functions to I&C systems and components. Due to the characteristics of computer-based technology, component failures can negatively affect several I&C functions, so that the reliability proof of the I&C systems requires the accomplishment of I&C system design analyses throughout the I&C life-cycle. On one hand, this paper proposes the restructuring of the sequential IEC 61513 I&C life-cycle according to the V-model, so as to adequately integrate the concept of verification and validation. On the other hand, based on a metamodel for the modeling of I&C systems, this paper introduces a method for the modeling and analysis of the effects with respect to the superposition of failure combinations and event sequences on the I&C system design, i.e. the temporal change of physical structure is analyzed. In the first step, the method is concerned with the modeling of the I&C systems. In the second step, the method considers the analysis of temporal change of physical structure, which integrates the concepts of the diversity and defense-in-depth analysis, fault tree analysis, event tree analysis, and failure mode and effects analysis.
DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE
Kim, S.K. ; Ko, W.I. ; Lee, Yoon Hee ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 665~674
DOI : 10.5516/NET.06.2012.081
This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.
EUTECTIC(LiCl-KCl) WASTE SALT TREATMENT BY SEQUENCIAL SEPARATION PROCESS
Cho, Yung-Zun ; Lee, Tae-Kyo ; Choi, Jung-Hun ; Eun, Hee-Chul ; Park, Hwan-Seo ; Park, Geun-Il ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 675~682
DOI : 10.5516/NET.06.2013.022
The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at
; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 - 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 - 86 % for Cs and Sr was estimated.
INTERACTION STUDIES OF CERAMIC VACUUM PLASMA SPRAYING FOR THE MELTING CRUCIBLE MATERIALS
Kim, Jong Hwan ; Kim, Hyung Tae ; Woo, Yoon Myung ; Kim, Ki Hwan ; Lee, Chan Bock ; Fielding, R.S. ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 683~688
DOI : 10.5516/NET.07.2013.012
Candidate coating materials for re-usable metallic nuclear fuel crucibles, TaC, TiC, ZrC,
, were plasmasprayed onto a niobium substrate. The microstructure of the plasma-sprayed coatings and thermal cycling behavior were characterized, and U-Zr melt interaction studies were carried out. The TaC and
coating layers had a uniform thickness, and high density with only a few small closed pores showing good consolidation, while the ZrC, TiC, and
coatings were not well consolidated with a considerable amount of porosity. Thermal cycling tests showed that the adhesion of the TiC, ZrC, and
coating layers with niobium was relatively weak compared to the TaC and
coatings. The TaC and
coatings had better cycling characteristics with no interconnected cracks. In the interaction studies, ZrC and
coated rods showed significant degradations after exposure to U-10 wt.% Zr melt at
for 15 min., but TaC, TiC, and
coatings showed good compatibility with U-Zr melt.
FABRICATION OF GD CONTAINING DUPLEX STAINLESS STEEL SHEET FOR NEUTRON ABSORBING STRUCTURAL MATERIALS
Choi, Yong ; Moon, Byung M. ; Sohn, Dong-Seong ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 689~694
DOI : 10.5516/NET.07.2013.015
A duplex stainless steel sheet with 1 wt.% gadolinium was fabricated for a neutron absorbing material with high strength, excellent corrosion resistance, and low cost as well as high neutron absorption capability. The microstructure of the as-cast specimen has typical duplex phases including 31% ferrite and 69% austenite. Main alloy elements like chromium (Cr), nickel (Ni), and gadolinium (Gd) are relatively uniformly distributed in the matrix. Gadolinium rich precipitates were present in the grains and at the grain boundaries. The solution treatment at
for 50 minutes followed by the hot-rolling above
after keeping the sheet at
for 1.5 hours are important points of the optimum condition to produce a 6 mm-thick plate without cracking.
EVALUATION OF BRACHYTHERAPY FACILITY SHIELDING STATUS IN KOREA OBTAINED FROM RADIATION SAFETY REPORTS
Keum, Mi Hyun ; Park, Sung Ho ; Ahn, Seung Do ; Cho, Woon-Kap ;
Nuclear Engineering and Technology, volume 45, issue 5, 2013, Pages 695~700
DOI : 10.5516/NET.08.2012.093
Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4%) included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 (
), as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.