• Title, Summary, Keyword: APR1400

Search Result 276, Processing Time 0.039 seconds

THE DESIGN FEATURES OF THE ADVANCED POWER REACTOR 1400

  • Lee, Sang-Seob;Kim, Sung-Hwan;Suh, Kune-Yull
    • Nuclear Engineering and Technology
    • /
    • v.41 no.8
    • /
    • pp.995-1004
    • /
    • 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.

Structural Analysis and Measuring Locations of Upper Guide Structure Assembly in APR1400 (APR1400 상부안내구조물 집합체 구조해석 및 측정위치)

  • Ko, Do-Young;Kim, Kyu-Hyung;Kim, Sung-Hwan
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
    • /
    • /
    • pp.306-311
    • /
    • 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.

  • PDF

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

  • Ko, Do-Young;Kim, Kyu-Hyung
    • Nuclear Engineering and Technology
    • /
    • v.45 no.2
    • /
    • pp.249-256
    • /
    • 2013
  • In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

Establishment of Response Instrumentation Test Acceptance Criteria for APR1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program (APR1400 원자로내부구조물 종합진동평가 응답측정시험 허용기준 수립)

  • Ko, Do-Young;Kim, Kyu-Hyung;Kim, Sung-Hwan
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
    • /
    • /
    • pp.212-218
    • /
    • 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.

  • PDF

Response Instrumentation Test Acceptance Criteria for APR1400 RVI CVAP (APR1400 원자로내부구조물 종합진동평가 응답측정시험 허용기준)

  • Ko, Do-Young;Kim, Kyu-Hyung;Kim, Sung-Hwan
    • Transactions of the Korean Society for Noise and Vibration Engineering
    • /
    • v.21 no.11
    • /
    • pp.1036-1042
    • /
    • 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.

Design Re-engineering of the Lower Support Structure of the APR1400 Reactor Internals

  • Tung, Nguyen Anh;Namgung, Ihn
    • Journal of the Korea Society of Systems Engineering
    • /
    • v.13 no.1
    • /
    • pp.25-31
    • /
    • 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.

Written Plan of CVAP Design Control Document for APR1400 U.S. Design Certification (APR1400 미국 설계인증을 위한 종합진동평가 심사서류 작성 방안)

  • Ko, Do Young;Kim, Dong Hak;Park, Young Sheop
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
    • /
    • /
    • pp.102-105
    • /
    • 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.

  • PDF

Development of Engineering Program for APR1400 Feedwater Supplying System (APR1400 급수공급계통 엔지니어링 프로그램 개발)

  • Yeom, Dong Un;Ju, Tae Young;Hyun, Jin Woo
    • Journal of Energy Engineering
    • /
    • v.26 no.2
    • /
    • pp.12-22
    • /
    • 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.

Comprehensive Vibration Assessment Program Measurement Test Plan for Advanced Power Reactor 1400 (신형경수로 1400 종합진동평가프로그램 측정시험 계획)

  • Ko, Do-Young;Kim, Kyu-Hyung
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
    • /
    • /
    • pp.589-595
    • /
    • 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.

  • PDF

Analysis of Fluid-Induced Vibration in the APR1400 Steam Generator Tube (신형경수로1400 증기발생기 전열관의 유체유발진동 해석)

  • 이광한;정대율;변성철
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
    • /
    • /
    • pp.84-91
    • /
    • 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.

  • PDF