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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 32, Issue 6 - Dec 2000
Volume 32, Issue 5 - Oct 2000
Volume 32, Issue 4 - Aug 2000
Volume 32, Issue 3 - Jun 2000
Volume 32, Issue 2 - Apr 2000
Volume 32, Issue 1 - Feb 2000
Selecting the target year
An Assessment of Reactor Vessel Integrity Under In-Vessel Vapor Explosion Loads
Bang, Kwang-Hyun ; Cho, Jong-Rae ; Park, Soo-Yong ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 299~308
A safety assessment of reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The core melt relocation parameters were chosen within the ranges of physically realizable bounds. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were peformed using ANSYS code. Then, the calculated strain results and the established failure criteria were used in determining the failure probability of the lower head, In the explosion analyses, it is shown that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations. Strain analyses show that the vapor explosion-induced lower head failure is not possible under the present framework of assessment. The result of static analysis using the conservative explosion-end pressure of 50 MPa also supports the conclusion. It is recommended, however, that an assessment of fracture mechanics for preexisting cracks be also considered to obtain a more concrete conclusion.
A Study on Enhancement of Np Extraction by TBP Through the Electrochemical Adjustment of Np Oxidation State by Using a Glassy Carbon Fiber Column Electrode
Kim, Kwang-Wook ; Song, Kee-Chan ; Lee, Eil-Hee ; Park, In-Kyu ; Yoo, Jae-Hyung ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 309~315
The changes of Np oxidation state in nitric acid and the effect of nitrous acid on the oxidation state were analyzed by spectrophotometry, solvent extraction, and electrochemical methods. An enhancement of Np extraction to 30 vol.% TBP was carried out through adjustment of Np oxidation state by using a glassy carbon fiber column electrode system. The information of electrolytic behavior of nitric acid was important because the nitrous acid affecting the Np redox reaction was generated during the electrolytic adjustment of the Np oxidation state. The Np solution used in this work consisted of Np（V) and Np（Ⅵ）without （IV）. The composition of Np（V) in the range of 0.5M -5.5 M nitric acid was 32% ~ 19%. The electrolytic oxidation of Np（V) to Np（Ⅵ）in the solution enhanced Np extraction efficiency about five times higher than the case without the electrolytic oxidation. It was confirmed that the nitrous acid of less than about 10-5 M acted as a catalyst to accelerate the chemical oxidation reaction of Np（V) to Np（Ⅵ）.
A Study on Design of the Trip Computer for ECC System Based on Dynamic Safety System
Kim, Seog-Nam ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 316~327
The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.
Modal Analysis of Coaxial Shells with Fluid-Filled Annulus
Jhung, Myung-Jo ; Kim, Yong-Beum ; Jeong, Kyeong-Hoon ; Park, Suhn ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 328~341
Investigated in this study are the modal characteristics of the coaxial cylindrical shells with fluid-filled annulus. Theoretical method is developed to find the natural frequencies of the shell using the finite Fourier series expansion, and their results are compared with those of finite element method to verify the validation of the method developed. The effect of the fluid-filled annulus and the boundary conditions on the modal characteristics of the coaxial shells is investigated using a finite element modeling.
An Analysis on Remediation of Soil Contaminted with Cobalt by Solvent Flushing
Kim, Gyenam ; Kyungsuk Suh ; Huijun Won ; Joonbo Shim ; Wonzin Oh ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 342~349
A soil whose texture is silt loam was collected for the study from an area around a nuclear facility in Korea. The equilibrium sorption coefficient between Co
in water and the soil was 1.51/kg, on the other hand, that between Co
in EDTA and the soil was 0.21/kg. The values calculated by the developed nonequilibrium sorption code corresponded to the experimental values better than those calculated by the existing equilibrium sorption code. When an EDTA solution was used as a solvent to decontaminate Co
in the soil column, the relative Co
concentrations of the effluent were higher at 2~10 pore volumes than those of the case using water. The soil in the column was decontaminated by 95.5% of the total amount of Co
after being flushed with EDTA solution of 20 pore volumes.e volumes.
Derived Limits for Radiological Protection Against ionizing Radiation Based on ICRP-60 Recommendations
Jang, Si-Young ; Lee, Byung-Soo ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 350~360
In Korea, the dose limits are reduced and are set at the ICRP-60 iimits. However, derived limits tabulated as MPC in air and water are still specified in Notice No.98-12. There are some discrepancies between the primary dose limits and MPCs in air and water. Therefore, in order to accept ICRP-60 recommendations fully, derived limits such as ALI, DAC, ECL for radiological protection against ionizing radiation based on ICRP-60 recommendations were calculated using modified methods of those of 10 CFR part 20, dose limits and committed effective dose coefficients of the Basic Safety Standards of the IAEA. The derived limits in this study were also compared with those prescribed in 10 CFR part 20 as well as MPCs of Notice No. 98-12 in order to analyze the impact of implementing derived limits on nuclear facilities. ECLs in air and water for the control of radioactive discharge into the environment in this study are shown to have lower values (i.e. more conservative), for most part, than those in Notice No. 98-12. Especially, for uranium elements, ECLs in water are approximately a magnitude in the order of two lower than those in Notice No.98-12.
Calculation of Proton-Induced Reactions on Tellurium Isotopes Below 60 MeV for Medical Radioisotope Production
Kim, Doohwan ; Jonghwa Chang ; Yinlu Han ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 361~371
The 123Te(p,n)123I, 124Te(p,n)124I and 124Te(p,2n)123I reactions, among the many reaction channels opened, are the major reactions under consideration from a diagnostic purpose because reaction residuals as the gamma emitters are used for most radiophamaceutical applications involving radioiodine. Based on the available experimental data, the absorption cross sections and elastic scattering angular distributions of the proton-induced nuclear reaction on Te isotopes below 60 MeV are calculated using the optical model code APMNK. The transmission coefficients of neutron, proton, deuteron, trition and alpha particles are calculated by CUNF code and are fed into the GNASH code. By adjusting level density parameters and the pair correction values of some reaction channels, as well as the composite nucleus state density constants of the pre-equilibrium model, the production cross sections and energy-angle correlated spectra of the secondary light particles, as well as production cross sections and energy distributions of heavy recoils and gamma rays are calculated by the statistical plus pre-equilibrium model code GNASH. The calculated results are analysed and compared with the experimental data taken from the EXFOR. The optimized global optical model parameters give overall agreement with the experimental data over both the entire energy range and all tellurium isotopes.
Effect of Alloying Elements on the Thermal Creep of Zirconium Alloys
Cheol Nam ; Kim, Kyeong-Ho ; Lee, Myung-Ho ; Jeong, Yong-Hwan ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 372~378
The effect of alloying elements on the creep resistance of Zr alloys was investigated using thermal creep tests that were performed as a part of advanced fuel cladding development. The creep tests were conducted at 40
and 150 MPa for 240 hr. A statistical model was derived from the relationship between the steady-state creep rate and the content of individual alloying elements. The creep strengthening effect decreased in the following sequence : Nb, Sn, Mn, Cr, Mo, Fe and Cu. The high creep resistance of Sn and the opposite effect of Fe on zirconium alloys seem to be associated with their lowering and enhancing, respectively, the self-diffusivity of the zirconium matrix.
Sensitivity Studies on Thermal Margin of Reactor Vessel Lower Head During a Core Melt Accident
Kim, Chan-Soo ; Kune Y. Suh ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 379~394
As an in-vessel retention (IVR) design concept in coping with a severe accident in the nuclear power plant during which time a considerable amount of core material may melt, external cooling of the reactor vessel has been suggested to protect the lower head from overheating due to relocated material from the core. The efficiency of the ex-vessel management may be estimated by the thermal margin defined as the ratio of the critical heat flux (CHF）to the actual heat flux from the reactor vessel. Principal factors affecting the thermal margin calculation are the amount of heat to be transferred downward from the molten pool, variation of heat flux with the angular position, and the amount of removable heat by external cooling In this paper a thorough literature survey is made and relevant models and correlations are critically reviewed and applied in terms of their capabilities and uncertainties in estimating the thermal margin to potential failure of the vessel on account of the CHF Results of the thermal margin calculation are statistically treated and the associated uncertainties are quantitatively evaluated to shed light on the issues requiring further attention and study in the near term. Our results indicated a higher thermal margin at the bottom than at the top of the vessel accounting for the natural convection within the hemispherical molten debris pool in the lower plenum. The information obtained from this study will serve as the backbone in identifying the maximum heat removal capability and limitations of the IVR technology called the Cerium Attack Syndrome Immunization Structures (COASISO) being developed for next generation reactors.
Sensitivity Analyses for Maximum Heat Removal from Debris in the Lower Head
Kim, Yong-Hoon ; Kune Y. Suh ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 395~409
Parametric studies were performed to assess the sensitivity in determining the maximum in-vessel heat removal capability from the core material relocated into the lower plenum of the reactor pressure vessel (RPV）during a core melt accident. A fraction of the sensible heat can be removed during the molten jet delivery from the core to the lower plenum, while the remaining sensible heat and the decay heat can be transported by rather complex mechanisms of the counter-current flow limitation (CCFL) and the critical heat flux (CHF）through the irregular, hemispherical gap that may be formed between the freezing oxidic debris and the overheated metallic RPV wall. It is shown that under the pressurized condition of 10MPa with the sensible heat loss being 50% for the reactors considered in this study, i.e. TMI-2, KORI-2 like, YGN-3&4 like and KNGR like reactors, the heat removal through the gap cooling mechanism was capable of ensuring the RPV integrity as much as 30% to 40% of the total core mass was relocated to the lower plenum. The sensitivity analysis indicated that the cooling rate of debris coupled with the sensible heat loss was a significant factor The newly proposed heat removal capability map (HRCM) clearly displays the critical factors in estimating the maximum heat removal from the debris in the lower plenum. This map can be used as a first-principle engineering tool to assess the RPV thermal integrity during a core melt accident. The predictive model also provided ith a reasonable explanation for the non-failure of the test vessel in the LAVA experiments performed at the Korea Atomic Energy Research Institute (KAERI), which apparently indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices.
Two Dimensional Analysis for the External Vessel Cooling Experiment
Yoon, Ho-Jun ; Kune Y. Suh ;
Nuclear Engineering and Technology, volume 32, issue 4, 2000, Pages 410~423
A two-dimensional numerical model is developed and applied to the LAVA-EXV tests performed at the Korea Atomic Energy Research Institute (KAERI) to investigate the external cooling effect on the thermal margin to failure of a reactor pressure vessel (RPV) during a severe accident. The computational program was written to predict the temperature profile of a two-dimensional spherical vessel segment accounting for the conjugate heat transfer mechanisms of conduction through the debris and the vessel, natural convection within the molten debris pool, and the possible ablation of the vessel wall in contact with the high temperature melt. Results of the sensitivity analysis and comparison with the LAVA-EXV test data indicated that the developed computational tool carries a high potential for simulating the thermal behavior of the RPV during a core melt relocation accident. It is concluded that the main factors affecting the RPV failure are the natural convection within the debris pool and the ablation of the metal vessel, The simplistic natural convection model adopted in the computational program partly made up for the absence of the mechanistic momentum consideration in this study. Uncertainties in the prediction will be reduced when the natural convection and ablation phenomena are more rigorously dealt with in the code, and if more accurate initial and time-dependent conditions are supplied from the test in terms of material composition and its associated thermophysical properties.