• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.306-311
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• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.11
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• pp.1036-1042
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• 2011

APR1400 RVI CVAP using the non-prototype category II is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during pre-operational and initial startup testing and to determine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the inner barrel assembly and the upper guide structure were established to 0.03 in and 0.01 in, respectively.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.212-218
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• 2011

APR1400 RVI CVAP using the non-prototype category is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during preoperational and initial startup testing and to detemine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the IBA and the UGS were established to 0.03 in and 0.01 in, respectively.

• Journal of the Korean Society of Systems Engineering
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• v.13 no.1
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• pp.25-31
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• 2017

This paper aims to evaluate the conservatism in the design of APR1400 (Advanced Pressurized water Reactor 1400 designed by KHNP) reactor internals component, the LSS (Lower Support Structure). Re-engineering of the LSS is done based on the system design condition data and applicable ASME code that was used for the original APR1400 design. Systems engineering approach is applied to design the LSS of APR1400 without refering APR1400 LSS dimensional parameters and tries to verify important design parameters of APR1400 LSS as well as the validity of the re-engineering design process as independent verification method of reactor component design. Systems engineering approach applied in this study following V-model approach. The re-engineered LSS design showed more than enough conservatism for static loading case. The maximum deflection of LSS is under 1mm (calculated value is 0.25mm) from 4000 mm diameter of LSS. Hence the deflection can be ignored in other reactor internals for structural integrity assessment. Especially the effect of LSS deflection on fuel assembly can be minimized and which is one of the main requirements of LSS design. It also showed that the maximum stress intensity is 2.36MPa for the allowable stress intensity of 60.1 MPa. The stress resulted from the static load is also very small compared to the maximum allowable stress intensity, hence there is more than enough conservatism in the LSS design.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Journal of Energy Engineering
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• v.26 no.2
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• pp.12-22
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• 2017

Korea Hydro & Nuclear Power Co. (KHNP) has implemented engineering programs for operating nuclear power plants. Engineering programs are maintenance rule (MR), functional importance determination (FID), single point vulnerability (SPV) and functional equipment group (FEG). Recently, KHNP has developed engineering programs for APR1400 feedwater supplying system to establish the advanced engineering system and will verify the suitability of engineering programs through implementing in new nuclear power plant. Consequently, it is expected that the reliability of APR1400 feedwater supplying system will be improved by implementing engineering programs.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.04a
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• pp.589-595
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being verified on the integrity of RVI for the design life of the plant by performing the non-prototype category-2 type on the US Nuclear Regulatory Commission Guide(NRC RG) 1.20, for which consists of a vibration and stress analysis program, a limited vibration measurement program, an inspection program, and the correlation of these programs. The aim of this paper is to describe the plan for the vibration measurement, test and acceptance criteria portion, and documentation and results of the APR1400 RVI CVAP. We will conduct the limited vibration measurement program of the APR1400 RVI CVAP according to the measurement plan and the vibration measurement testing in this paper.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.22 no.5
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.