• Journal of the Korean Society of Marine Environment & Safety
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• v.27 no.2
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• pp.355-362
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• 2021

To facilitate low-pressure selective catalyst reduction (L.P SCR), a high exhaust-gas temperature of a four-stroke diesel engine for a ship's generator is required. This study aimed at reducing the exhaust-gas temperature by adjusting the valve open-close timing and fuel injection timing to satisfy the operating conditions of L.P SCR and prevent accidents associated with the generator engine due to high temperature. To lower exhaust-gas temperature, the angle of the camshaft was adjusted and the shim of the fuel injection pump was added. As a result, the maximum explosion pressure increased and the average of the turbocharger outlet temperature dropped. Considering the heat loss from the turbocharger outlet to the SCR inlet, the operation condition for L.P SCR was satisfied with 290 ℃. The study demonstrates that safe operation of a diesel generator can be achieved by lowering the exhaust-gas temperature.

• Journal of the Korean Electrochemical Society
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• v.10 no.4
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• pp.301-305
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• 2007

The performance of polymer electrolyte membrane fuel cell could be decreased due to coolant leaked from connection part. Micro pump was used to put small amount of coolant and investigate the effect on fuel cell. The stoichiometric ratio of hydrogen/air was 1.5/2.0, both side of gas was fully humidified, and current density of $400mA/cm^2$ was used as standard condition in this experiment. Constant current method was used to check performance recovery from coolant effect in 3 cell stack. The performance was recovered when coolant was injected in cathode side. On the other hand, the performance was not recovered when coolant was injected in anode side. Ethylene glycol could be converted to CO in oxidation process and cause poisoning effect on platinum catalyst or be adhered on GDL and cause gas diffusion block effect resulting performance decrease. Water with nitrogen gas was supplied in anode side to check performance recovery. Polarization curve, cyclic voltammetry, electrochemical impedance spectroscopy was used to check performance, and gas chromatography was used to check coolant concentration. Constant current method was not enough in full recovery of performance. However, water injection method was proved good method in full recovery of performance.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.11
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• pp.1371-1379
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• 2012

This paper presents a method for estimating the contribution of vibration sources in gasoline direct injection engine parts with a multiple-input system. A partial coherence function was used to identify the cause of the linear dependence indicated by an ordinary coherence function. To apply the partial coherence function to vibration source identification in the powertrain system of a gasoline direct injection engine, a virtual model of a two-input and single-output system is simulated. For the validation of this model, the vibration of the powertrain parts was measured by using triaxial accelerometers attached to the selected vibration sources-a high-pressure pump, fuel rail, injector, and pressure sensor. After calculating the partial coherence between each source based on the virtual model, the vibration contribution of the powertrain system is calculated. This virtual model based on the partial coherence function is implemented to determine the quantitative vibration contribution of each powertrain part.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.