• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.555-561
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• 1995

This paper examines the applicability of the modified Borresen's coarse-mesh method(MBSN) to the core power distribution monitoring program development for the Yonggwang nuclear power plant unit 3(YGN 3) which uses fixed incore detectors for monitoring core power distribution. In so doing the modified Borresen's coarse-mesh equations are solved with core internal boundary conditions provided by the fixed incore detectors and three-dimensional core power distributions are com puted for the first-cycle core of the YGN 3 PWR. The results are compared with predictions of the COLSS(Core Operating Limit Supervisory System) which is the axial power shape monitoring program of the YGN 3. It is shown that the modified Borresen's method can reproduce the core axial power shape more closely than the COLSS. Because of other advantages in computing speed and predictive capability, n conclude that the proposed MBSN has a promising practical application for core power distribution monitoring program development.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.228-237
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• 2019

The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at $325^{\circ}C$ simulating the normal operating conditions of a nuclear power plant. The as-received Alloy 600 showed crack growth rates (CGRs) between $9.6{\times}10^{-9}mm/s$ and $3.8{\times}10^{-8}mm/s$, and the as-welded Alloy 182 had CGRs between $7.9{\times}10^{-8}mm/s$ and $7.5{\times}10^{-7}mm/s$ within the range of the applied loadings. These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be $2.8{\times}10^{-9}mm/s$. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure. The cracking mode of Alloys 600 and 182 was an intergranular cracking; however, Alloy 152 was revealed to have a mixed (intergranular + transgranular) cracking mode. It appears that the Cr concentration and the microstructural features significantly affect the cracking resistance and the cracking behavior of Ni-base alloys in PWR primary water.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.10
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• pp.873-881
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• 2008

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.371-379
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• 1990

The Xenon spatial oscillation may give rise to operational difficulties in a nuclear power plant. In this study, in order to investigate the Xenon instability for a PWR, the frequency-domain technique is adopted by using Generalized Nyquist Criterion, which is more general and suitable for the multi-input/multi-output system. Also linearized modal fluxes are obtained by a modal expansion. This model has been implemented to test the axial Xenon stability of YGN-1 unit against the changes in plant operating parameters ; power level, control rod position, and core average burnup. The results show that the increase of power level and the deeper insertion of control rod have the destabilizing effect, and that the burnup progress makes the core less stable. Also the results show that the overestimation due to modal interaction was found not to be significant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.289-295
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• 2016

Kori Unit 1, the first PWR (Pressurized Water Reactor) plant in Korea, began its commercial operation in 1978 and will permanently shut down on June 18, 2017. After moving the spent fuels to SFP (Spent Fuel Pool) system, Kori Unit 1 will perform a full system chemical decontamination to reduce radiation levels inside the various plant systems. This paper will describe the operation concept of the full system chemical decontamination for Kori Unit 1 based on experiences overseas.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.87-111
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• 2021

Nuclear power plants (NPPs) produce radioactive waste and decommissioning this waste entails additional cost; determining these costs for various types and specifications of radioactive waste can be challenging. The purpose of this study is to identify major determinants of the decommissioning cost and their impact on NPPs. To this end, data from defunct NPPs were gathered and 2SLS (Two Stage Least Squares) regression models were developed to investigate the major contributors depending on the reactor types, viz. PWR (Pressurized Water Reactors) and BWR (Boiling Water Reactors). Additionally, cost estimations and the Monte Carlo simulation were performed as part of performance validation. Our study established that the decommissioning costs are primarily influenced by the level of radioactivity in the decommissioned waste, which can be realized from operational factors like operation period, overall efficiency, and plant capacity, as well as from duration of decommissioning and labour cost. While our study provides an improved statistical approach to recognize these factors, we acknowledge that our models have limitations in forecasting accurately which we envisage to bolster in future studies by identifying more substantive factors.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4408-4425
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• 2023

Alarm flood due to abnormality propagation is the most difficult alarm overloading problem in nuclear power plants (NPPs). Root-cause analysis is suggested to help operators in understand emergency events and plant status. Multilevel Flow Modeling (MFM) has been extensively applied in alarm management by virtue of the capability of explaining causal dependencies among alarms. However, there has never been a technique that can identify the actual root cause for complex alarm situations. This paper presents an automated root-cause analysis system based on MFM. The causal reasoning algorithm is first applied to identify several possible root causes that can lead to massive alarms. A novel root-cause ranking algorithm can subsequently be used to isolate the most likely faults from the other root-cause candidates. The proposed method is validated on a pressurized water reactor (PWR) simulator at HAMMLAB. The results show that the actual root cause is accurately identified for every tested operating scenario. The automation of root-cause identification and ranking affords the opportunity of real-time alarm analysis. It is believed that the study can further improve the situation awareness of operators in the alarm flooding situation.

• Journal of Radiation Protection and Research
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• v.17 no.2
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• pp.61-70
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• 1992

Photon, neutron and beta radiation fields were measured at PWR plants which are the representative types of nuclear power plant operated in Korea. The photon energy spectra were measured at locations in the auxiliary building during operation period and in the containment vessel(C/V) during shutdown period using a portable gamma spectrometer with a HPGe detector. The distribution of average energy was found to range from 440 to 780 keV in the C/V and from 280 keV to 760 keV in the auxiliary building, respectively. The average neutron energy measured at the five locations around the operation deck in the C/V in operation using a BMSS (Bonner Multi-Sphere Spectrometer) ranged from 20 keV to 210 keV. A computer code, BUNKI was used to unfold the spectrum. The beta energy spectra in the C/V and in the auxiliary building in annual outage were determined using 14 smear samples taken from the highly contaminated areas. The analysis showed that the representative corrosion product, $^{60}Co$ made main contribution to the beta energy field.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.611-616
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• 1997

The Level 1 probabilistic safety assessment(PSA) for Wolsong(WS) 2/3/4 nuclear power plant(NPPs) in design stage is performed using the methodologies being equivalent to PWR PSA. Accident sequence evaluation program(ASEP) human reliability analysis(HRA) procedure and technique for human error rate prediction(THERP) are used in HRA of WS 2/3/4 NPPs PSA. The purpose of this paper is to introduce the procedure and methodology of HRA in WS 2/3/4 NPPs PSA. Also, this paper describes the interim results of importance analysis for human actions modeled in WS 2/3/4 PSA and the findings and recommendations of administrative control of secondary control area from the view of human factors.