• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.389-394
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• 1996

The conceptual design of Korea Next Generation Reactor (KNGR), which is 3914 MWt PWR, includes the safety depressurization system (SDS) to comply with U.S. NRC's severe accident policy. In this analysis, it is assumed that three Monobloc Sebim valves are adopted for the SDS bleed valves of KNGR. The characteristic of Monobloc Sebim are modeled in the CE-FLASH-4AS/REM code for this analysis. The various feed and bleed (F&B) procedures with Sebim valves are investigated for total loss of feedwater (TLOFW) event. It is found that if operators open two out of three Sebim valves in conjunction with four HPSI pumps before hot leg temperature reaches saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. Therefore, this F&B procedure can be used for mitigating the TLOFW event of the KNGR. This result also demonstrates the feasibility of adopting the Monobloc Sebim valves for the SDS of KNGR.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.417-422
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• 1996

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECC design can tolerate a cold leg break of up to 10 inches with no core uncovery. However. since DVI line break with 6 inch diameter undergoes slight core uncovery. further investigation is required for KNGR SIS optimization.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.218-228
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• 1997

Correlations between the amount of DAW (Dry Active Waste) generated from present Korean PWRs and their operating parameters were analyzed. As the result of multi-variable linear regressions, a model predicting the volume of DAW using the number of shutdowns ( $f_{FS}$ ) and total personnel exposure ( $P_{\varepsilon}$) was derived. Considering one standard error bound, the model could successfully simulate about 8575 of the real data. In order to predict the amount of DAW to be generated from a KNGR another model was derived by taking into account the additional volume reduction by supercompaction system. In addition, the volume of WAW (Wet Active Waste) to be generated from KNGR (Korean Next Generation Reactor) was calculated by considering conceptual design data and replacement effect of radwaste evaporator with selective ion exchangers. Finally, total volume of SRW (Solid Radioactive Waste) to be generated from KNGR was predicted by inserting design goal values of $f_{FS}$ and $P_{\varepsilon}$ into the model. The result showed that the expected amount of SRW to be generated from KNGR would be in the range of 33~44㎥. $y^{-1}$ . It was proved that the value would meet the operational target of KNGR proposed by KEPCO, that is, 50㎥. $y^{-1}$ .

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.355-360
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• 1996

Human factors principles and evaluation methodologies are applied to design the MMIS of the KNGR. Human factors issues identified from the previous MMIS design of a nuclear power plant are considered in the development of the HFEPP. To manage human factors issues in the MMIS design of the KNGR, a conceptual Human Factors Issue Tracking System (HFITS) is also considered.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.663-668
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• 1997

한국형 차세대 원자로(KNGR)는 안전주입계통에 Advanced Design features를 채택하고 있는데, 그 중의 하나가 안전주입의 주입구를 Downcomer Annulus의 상부에 위치시킨 Direct Vessel Injection(DVI)으로서 영광 및 울진 3&4호기의 Cold Leg Injection(CLI)과는 다른 설계 개념이다. 본 논문에서는 DVI가 채택된 KNGR에 대하여 기존의 C-E형 발전소 해석에 적용한 C-E Evaluation Model(EM)을 사용하여 대형파단 냉각재상실사고를 해석해 보고자 하였다. 먼저 DVI의 Modeling은 KNOGR의 참조 발전소라 할 수 있는 System80+에서 Modeling한 것과 같이 CLI 해석에 사용한 Nodalization Scheme 중 Cold Leg Node에 연결된 SIT 만을 Downcomer Annulus Node에 연결하는 방법을 사용하여 DVI 해석을 수행하였다. 아울러 기존의 안전주입 형태인 CLI에 대한 해석을 KNGR에 대해 병행하여 수행함으로써 DVI와 CLI의 ECCS performance를 비교하고 CLI 대비 DVI의 특성을 알아보았다. 또한 DVI의 해석에 있어서 SIT와 Cold Leg이 함께 연결되는 Downcomer Annulus Node를 상하 2개로 분리하여 SIT와 Cold Leg 각각에 연결시킴으로써 DVI 주입구의 위치에 대한 보다 정확한 Modeling을 시도하였다. 그 결과 DVI 주입구의 높이를 고려한 경우가 DVI의 일반적 물리 현상에 근접하게 계산되는 것으로 판단되나 현재로서는 특별한 검증 수단이 없으므로 향후 Licensing 해석 수행에 앞서 방법론을 포함한 이에 대한 보다 심도 있는 검토가 필요할 것으로 판단된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.862-868
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• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.395-400
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• 1996

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are peformed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSI pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SIT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• Journal of Korean Institute of Industrial Engineers
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• v.28 no.2
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• pp.178-186
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• 2002

The objective of this study is to evaluate impact of computer-based man-machine interfaces of Korea Next Generation Reactor (KNGR) on the operator mental workload. Empirical experiments were conducted to measure the operator mental workloads of KNGR and Yong-Gwang Unit 3 and 4, respectively. A comparison analysis based on a NASA TLX revealed that Yong-Gwang Unit 3 and 4 were superior to KNGR in terms of the mental workload. Post-hoc analyses showed that the mental workload of senior reactor operators was significantly higher than those of reactor and turbine operators, regardless of plant types. The implications of the findings were discussed in detail.

• Proceedings of the KIEE Conference
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• 1999.11c
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• pp.600-602
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• 1999

원자력발전 비율이 증가함에 따라 전기에너지의 효율적인 이용과 원자력발전의 경쟁력 제고를 위하여 원자력발전소(이하 원전)의 부하추종운전 필요성이 점점 커지고 있다. G7과제의 하나로 개발되고 있는 차세대원자로(KNGR, Korean Next Generation Reactor)는 경쟁력 있는 원전의 설계를 위하여 "일일부하추종운전 능력의 확보"를 기본 성능요건의 하나로 하여 개발되고 있다. 그러나 수동으로 원자로출력분포를 제어하는 기존 원전의 제어방식으로는 상기목표를 충족시키기 어려워 원자로의 출력분포와 출력을 동시에 제어하는 새로운 자동 제어방식을 도입하였다. 본 논문에 기술된 제어방법은 원자로 출력분포 상태에 따른 비선형 제어방법이 적용되며 목표출력 부근에서의 Oscillatory Behavior 방지를 위해 설정된 Deadband 내에서의 다른 상태변수를 제어하기 위한 알고리즘도 포함된다. 개발된 제어방법의 성능을 확인하기 위해 원자로 증기공급계통 전체를 모델링한 성능분석 Simulator를 이용한 Numerical Simulation을 수행하였다. 일일부하추종운전은 100-50-100%P[$(10{\sim}16)-2-(10{\sim}4)-2$ hr] power cycle over a 24-hour period, 주파수제어는 일반적인 Grid Follow에 대해 Simulation하였다.