• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1765-1775
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• 2019

Multiple spurious operations (MSOs) mean multiple fire induced circuit faults causing an undesired operation of one or more systems or components. The Nuclear Energy Institute (NEI) of the United States published NEI 00-01 as guidelines for solving MSOs. And this guideline includes MSO scenarios of pressurized water reactor (PWR) and boiling water reactor (BWR). Nuclear power plant operators in U.S. analyzed MSOs under MSO scenarios included in NEI 00-01 and operators of PWRs in Korea also analyzed MSOs under the scenarios of NEI 00-01. As there are no pressurized heavy water reactors (PHWRs) in the United States, MSO scenarios of PHWRs are not included in the NEI 00-01 and any feasible scenarios have not been developed. This paper developed MSO scenarios which can be applied to PHWRs by reviewing the 63 MSO scenarios included in NEI 00-01. This study found that seven scenarios out of the 63 MSO scenarios can be applied and three more scenarios need to be developed.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• Proceedings of the KIEE Conference
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• 1984.07a
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• pp.92-95
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• 1984

This paper describes a general method for the design of variable structure Model-Following Control systems (VSMFC). This design concept is developed using the theory of variable structure systems and slide mode. The feasibility and the advantages of the method are illustrated by applying it to a 1000 MWe Boiling Water Reactor. The control is studied in the range of 85 - 90 % of rated power for load-following control. A set of 12 nonlinear differential eq. are used to simulate the total plant. A 6th order linear model has been developed from these equations at 85% of rated power. The obtained controller is shown by simulations to be able to compensate for a plant parameter variation over a wide power range.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Textile Coloration and Finishing
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• v.7 no.2
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• pp.55-62
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• 1995

Plasma polymerization of Perfiuoropropene(PFP) and n-Hexane was carried out in a tubular type reactor by means of 13.56MHz radio frequency generator at the fixed RF discharge power of 25W and at the pressures of 100mTorr, 140mTorr and 200mTorr. The thin films were deposited on PAN fabrics in order to improve the dimemsional stability of woven states in hot water laundry. IR spectroscopy was used for the analysis of the structures of the thin films deposited and SEM for examination of surfaces of the fabrics. the PAN fabrics, which were coated by thin films at several experimental conditions, were immersed in boiling water for 2 hours and then the dimension stability of woven states were evaluated. In spite of very thin films, the results of surface modification were satisfactory. In general the performace of thin films by PFP was superior to that of n-Hexane.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.255-268
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• 2008

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.

• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.11-17
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• 1984

Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA ($\leq$0.05 ft$^2$). For a water-side break with the break size larger than 0.006 ft$^2$, fluid-loss through break exceeds the makeup. If the break size is larger than 0.008ft$^2$, decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 900 MWe or 1200 MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervention was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1321-1329
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• 2016

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

• Journal of Advanced Marine Engineering and Technology
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• v.10 no.1
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• pp.74-84
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• 1986

Ejectors, having no moving, lubricating and leaking parats, are widely used as marine pumps because of its high working confidence. For instance, uses in ships are stripping in crude oil tank, bilge discharge in engine room, ballast water pumping on are carrier, and brine discharge from fresh water generator. And it is also used as cooling water recirculating pump in boiling water type nuclear reactor and deep-well pump. It is not easy to determine the optimal dimension for designing each ejector agreed with its suggested performance condition, because complicated calculations must be repeated to obtain the maximum efficiency affected by flowrate ratio, head ratio, area ratio and so on. Therefore, it is considered that the CAD (Computer-Aided Design) for ejector is a powerful method for design according to the individual design condition. In this paper, a computer program for water ejector design is developed based on the previous paper on theoretical analysis and experimental results for water ejector. And from the theoretical approach, an equation for the working limit and an equation for determing the shape of throat are obtained. The validity of the present computer program is sufficiently confirmed through the comparison of the computed results with the main dimension of the previous manufactured water ejector. This program will be easily developed as the CAD for various kinds of ejectors, including steam ejector.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.544-554
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• 1995

Pool boiling of dilute aqueous solutions of polyethylene oxide polymer has been experimentally investigated for the purpose of understanding the physical mechanisms of the suppression of vapor explosions in this polymer solution. Tn solid spheres of 22.2mm and 9.5mm-diameter ore heat-ed and quenched in the polymer solutions of various concentrations at 3$0^{\circ}C$. The results showed that minimum film boiling temperature($\Delta$ $T_{MFB}$) in this highly-subcooled liquid rapidly decreased from over $700^{\circ}C$ for pure water to about 15$0^{\circ}C$ as the polymer concentration was increased up to 300ppm for 22.2mm sphere, and it decreased to 35$0^{\circ}C$ for 9.5mm sphere. This large decrease of minimum film boiling temperature in this aqueous polymer solution may explain its ability to suppress spontaneous vapor explosions. Also, tests with applying a pressure wave showed that the vapor film behaved more stable against an external disturbance at higher polymer concentrations. These observations together with the experimental evidences of vapor explosion suppression in dilute polymer solutions suggest that the application of polymeric additives such as polyethylene oxide as low as 300ppm to reactor emergency coolant be considered to prevent or mitigate energetic fuel-coolant interactions during severe reactor accidents.s.