• Journal of the Korean Society of Safety
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• v.35 no.5
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• pp.115-120
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• 2020

From 2009 when the CNS facility was installed, the number of reactor failures due to abnormal CNS facility system has increased significantly. Of the total of 19 nuclear reactor shutdowns over the six years from 2009 to 2019, there were 10 nuclear reactor shutdowns associated with the CNS facility, which are very numerous. Therefore, this report intends to analyze the history of nuclear reactor shutdowns due to CNS facility system failure in detail, and to present the root cause and solution to problems. As a result of FMEA implementation of CNS facility system, a total of 76 SPVs were selected. In addition, 10 cases of reactor shutdown history due to CNS facility system abnormalities were analyzed in detailed, and improvement plans for solving the root cause and problem were suggested for each trip history. The results of this study are expected to be able to operate the domestic research reactor and CNS facilities more stably by providing effective measures to prevent recurrence of CNS facilities and reactor trips.

• Proceedings of the KIEE Conference
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• 2005.11b
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• pp.110-112
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• 2005

Recently, the reactor trip in nuclear power plants has mostly occurred by the failure of secondary electrical facilities. Some kinds of technical standards or acceptable criteria are required to regulate those facilities. Therefore, in this paper the failures of secondary electrical facilities which cause reactor trip are analyzed, and the inspection technologies including guidelines for secondary electrical facilities are developed.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Journal of the Korean Society of Safety
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• v.14 no.1
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• pp.177-184
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• 1999

PSA(Probabilistic Safety Assessment) methodology is widely used on assessing the safety of Nuclear Power Plants(NPPs) quantitatively in the domestic nuclear field. Initiating event frequencies are absolutely needed to conduct PSA, and they considerably affect PSA results. There is no domestic database where domestic trip event cases are reflected, so they are used to assess the safety of NPPs that are from the foreign database. In this paper, operating experience data from the Korean NPPs was collected and analyzed for the trip event cases, which are necessary to determine the initiating events and their frequencies. Korean NPPs have experienced five of 16 initiating events, which we LOFW. LOCV, LOCCW, LOOP and GTRN as a result of analyzing the trip event cases. Initiating frequencies based on the domestic trip event cases are analyzed, and they are similar to that from the foreign database.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.5-6
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• 2005

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.23-29
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• 2016

The Second Shutdown Drive Mechanism (SSDM) provides an alternate and independent means of reactor shutdown. The Second Shutdown Rods (SSRs) of SSDMs are poised at the top of the core by the hydraulic force driven from a hydraulic system during normal operation. The rods drop by gravity when a trip is commended by a Reactor Protection System, Alternate Protection System, Automatic Seismic Trip System or operator by means of power off solenoid valves of hydraulic system. This paper describes the test results of seismic qualification of a prototype SSDM hydraulic system to demonstrate that its structural integrity and operability (functionality) are maintained during and after seismic excitations, that is, an adequacy of the SSDM design. From the results, this paper shows that the SSDM hydraulic system satisfies all its design requirements without any malfunctions during and after seismic excitations.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.

• The KSFM Journal of Fluid Machinery
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• v.10 no.2 s.41
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• pp.41-45
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• 2007

The shutoff unit was designed to provide rapid insertion of neutron absorbing material into the reactor core to shutdown the reactor quickly and also to withdraw the absorber slowly to avoid a log-rate trip. Four shutoff units were installed on the HANARO reactor but the half-core test facility was equipped with one shutoff unit. The reactor trip or shutdown is accomplished by four shutoff units by insertion of the shutoff rods. The shutoff rod(SOR) is actuated by a directly linked hydraulic cylinder on the reactor chimney, which is pressurized by a hydraulic pump. The rod is released to drop by gravity, when triplicate solenoid valves are de-energized to vent the cylinder. The hydraulic pump, pipe and air supply system are provided to be similar with the HANARO reactor. The shutoff rod drops for 647mm stroke within 1.13 seconds to shut down the reactor and it is slowly inserted to the full down position, 700mm, with a damping. We have conducted the drop test of the shutoff rod in order to show the performance and the structural integrity of operating system of the shutoff unit. The present paper deals with the 647mm drop time and the withdrawal time according to variation of the pool water temperature, the water level and the core flow.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.41-48
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• 2012

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.