• 제목/요약/키워드: Small PWR

검색결과 64건 처리시간 0.02초

무붕산 알칼리 냉각재 온도 증가에 따른 Type 630 스테인리스강의 부식특성 평가 연구 (A Study on Accelerated Corrosion Rate of Stainless Steel Type 630 with Increasing Temperature of B-free Alkaline Coolant)

  • 박정수;임상엽;전순혁;김주성;오정목;심희상
    • 한국압력기기공학회 논문집
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    • 제20권1호
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    • pp.49-55
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    • 2024
  • Stainless 630 (or 17-4PH) is a precipitation-hardening martensitic stainless steel that has excellent mechanical properties and corrosion resistance. These characteristics make the STS630 to be used as a consisting material for various components such as spider, pin, spring, and spring retainer, of the control rod drive mechanism (CRDM) in pressurized water reactors (PWRs). In general, it is well known that the oxide layer of stainless steel consists of a duplex layer, a compact inner layer of FeCr2O4 spinel, and a coarse-grained outer layer of Fe3O4 spinel in PWR primary coolant condition. However, the characteristics of the oxide layer can be sensitively influenced by various water chemistry conditions such as temperature, dissolved oxygen, dissolved hydrogen, pH, pH adjuster type, and exposure time. In this work, we investigate the corrosion properties of the STS630 as a function of coolant temperature in an NH3 alkaline solution for its boron-free application in a small modular reactor, to confirm the feasibility for usage as a boron-free SMR structural material. As a result, oxide layer of corroded STS630 is consist of double-layer oxides consisting of a Cr-rich dense inner oxide and a Fe-rich polyhedral outer particles like as that in commercial PWR primary coolant. The corrosion rate of STS630 increases with increase in test time and temperature and the corrosion rate-time model equation was developed based on experimental data. Overall, it is expected that the results in this study provides useful data for the corrosion behavior of STS630 in alkaline environments, contributing to the development of selecting suitable materials for SMRs.

채널간 교차류가 냉각재상실사고에 미치는 영향분석 (Analysis of Inter-channel Cross Flow Effect on PWR LOCA)

  • Park, Jong-Ho;Lee, Sang-Yong;Han, Ki-In
    • Nuclear Engineering and Technology
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    • 제20권2호
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    • pp.80-87
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    • 1988
  • 소형냉각재상실사고(SBLOCA)와 대형냉각재상실사고(LBLOCA)중에 노심의 Average Channel과 Hot Channel에서의 유량분포를 예측하였다. 아울러 REALP5/MOD2 코드를 사용하여 두 채널사이의 교차류고려여부가 실제사고 분석결과에 미치는 영향을 평가하였다. 현재까지 SBLOCA계산에서는 노심을 한개의 채널로 모델하는 것이 충분하다고 판단되어 왔으나 본 계산결과에 의하면 보수적인 계산을 위해서는 Hot Channel 모델링이 필요한 것으로 밝혀졌다. 그러나 LBLOCA Blowdown Phase존에서는 교차류의 고려 여부에 상관없이 Hot Channel 이 Average Channel보다 보수적 인 결과를 가져오며, 교차류의 영향도 미세한 것으로 판명되었다.

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가압경수로 부분충수 운전중 잔열제거 (RHR)계통 상실시 가압기 통로를 통한 배출유로 특성 분석 (Analysis of the Vent Path Through the Pressurizer Manway Under the Loss of Residual Heat Removal(RHR) System During Mid-Loop Operation in PWR)

  • 하귀석;김원석;장원표;류건중
    • Nuclear Engineering and Technology
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    • 제27권6호
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    • pp.859-869
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    • 1995
  • 본 연구는 가압경수로의 부분충수 운전중 잔열제거기능 상실사고 해석시 신뢰성을 확보하기 위해 RELAP5/MOD3.1 코드로 관련 대형 실험을 모의 계산하여, 사고시 예상되는 주요 물리적 현상의 파악과 코드의 예측능력을 평가하는 것이다. 대상 실험으로 선택된 BETHSY Test 6.9a는 이 사고중 증기발생기가 작동하지 않고, 가압기 Manway를 개방한 상태 (Configuration)를 모의한 실험이다. 이 연구 결과는 실제 원전 사고시 예상되는 중요 현상 뿐 아니라, 이에 영향을 미치는 민감한 인자를 파악하여 사고 해석결과의 유효성을 판단하는 데 상당히 기여할 것으로 기대한다. 연구결과 RELAP5/MOD3.1 코드는 대체적으로 계통의 과도기 거동은 타당하게 예측하고 있지만, 모의계산에서 Time-Step이 아주 짧아 막대한 시간이 소요된다는 문제점이 발견되었다. 그 외에도 노심팽창수위 (swollen level)를 과대평가하여 가압기의 수위 및 계통의 압력을 높게 계산하였다. 이로 인해 가압기를 통한 방출량도 과대계산하여 노심노출을 약 400초 빨리 예측하였다. 실험과 코드 예측결과를 종합할 때 가압기 Manway 만의 개방으로는 계통압력이 상승하고, 중력주입냉각수로는 노심수위 회복에 불충분하며, 결국 강제주입에 의해서 노심수위가 회복될 수 있음이 입증되었다.

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A COMPARATIVE OVERVIEW OF THERMAL HYDRAULIC CHARACTERISTICS OF INTEGRATED PRIMARY SYSTEM NUCLEAR REACTORS

  • NINOKATA HISASHI
    • Nuclear Engineering and Technology
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    • 제38권1호
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    • pp.33-44
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    • 2006
  • This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

Axial BP Zoning for the Soluble Boron Free Operation in Medium-Sized PWR

  • Kim, Jong-Chae;Kim, Myung-Hyun
    • 한국원자력학회:학술대회논문집
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    • 한국원자력학회 1996년도 추계학술발표회논문집(1)
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    • pp.59-64
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    • 1996
  • Feasibility of soluble boron free operation for the medium-sized commercial reactors was investigated. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with gadolinia burnable poison-high Gd enrichment and axial poison zoning. CASMO and NECTA-C code system checked axial offset and peaking factors as fuels burned up. A core with complex axial burnable poison zoning satisfied design goals - small excess reactivity for 18 month cycle. Therefore, critical bank positioning for three control rod banks was sought with ease. A.O. value and Fq value were kept within the safety limit.

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Fretting Wear Mechanisms of Zircaloy-4 and Inconel 600 Contact in Air

  • Kim, Tae-Hyung;Kim, Seock-Sam
    • Journal of Mechanical Science and Technology
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    • 제15권9호
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    • pp.1274-1280
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    • 2001
  • The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

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Fabrication and Evaluation of a New High-Temperature pH Sensor for Use in PWR Nuclear Power Plants

  • Jung, Yong-Ju;Yeon, Jei-Won
    • Bulletin of the Korean Chemical Society
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    • 제31권10호
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    • pp.2939-2942
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    • 2010
  • A new high-temperature pH sensor has been successfully developed by reforming the internal reference systems of the pH sensors based on oxygen-ion conducting ceramic membrane. The conventional internal reference system, a mixture of Ni and NiO, has been replaced with partially oxidized Ni powders, where Ni and NiO coexist on the surface of particles, in order to avoid the cumbersome mixing step of Ni and NiO particles. The partially oxidized Ni particles were made by oxidizing Ni under air atmosphere at $600^{\circ}C$ and characterized by X-ray diffraction (XRD) and FTIR spectroscopy. The viability of the pH sensor developed was assessed in boric acid (1000 ppm-B)/ lithium hydroxide (1 to 3 ppm-Li) buffer solutions at $280^{\circ}C$. The pH sensor showed excellent accuracy with a small error less than ${\pm}0.2$ pH units.

직무 네트워크 모형을 이용한 원자력발전소 제어실 운전원들의 수행도분석 (Performance analysis of operators in a nuclear power plant control room using a task network model)

  • 서상문;천세우;이용희
    • 대한인간공학회:학술대회논문집
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    • 대한인간공학회 1993년도 추계학술대회논문집
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    • pp.21-30
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    • 1993
  • This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

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Mechanical Integrity Evaluation on the Degraded Cladding Tube of Spent Nuclear Fuel Under Axial and Bending Loads During Transportation

  • Lee, Seong-Ki;Lee, Dong-Hyo;Park, Joon-Kyoo;Kim, Jae-Hoon
    • 방사성폐기물학회지
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    • 제19권4호
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    • pp.491-501
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    • 2021
  • This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

직접용기주입에 따른 유체혼합에 관한 연구 (An Investigation of Fluid Mixing with Direct Vessel Injection)

  • Cha, Jong-Hee;Jun, Hyung-Gil
    • Nuclear Engineering and Technology
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    • 제26권1호
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    • pp.63-77
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    • 1994
  • 이 연구는 가압경수로의 원자로 다운커머내에서 과도냉각시 직접용기주입에 따른 유체혼합현상을 가압열충격의 견지에서 시험모델을 사용하여 조사한 것이다. 시험모델은 ABB-CE System80+ 원자로 구조에 근거하여 설계되었다. 이 원자로에 대한 가능성 있는 가압열충격 사고로서 콜드레그 소형파단 냉각재 상실사고와 주중기관 판단 사고가 선정되었다. 시험은 두 부분으로 구성되는데 첫째 부분은 원자로 다운커머에서 직접용기 주입수와 기존냉각재간의 유체혼합을 가시화법에 의하여 시험한 것이고, 둘째 부분은 별도의 시험모델에서 직접용기주입에 따른 열적혼합을 시험한 것이다. 가시화 시험에서는 과도적 냉각기간중 직접용기 주입수와 1차 냉각재간의 물리적 상호작용이 밝혀졌다. 열적혼합시험에서는 소형파단 냉각재 상실사고시 직접용기주입에 의한 심한 냉각현상이 다운커머내서 관찰되었다. 측정된 온도곡선은 소형파단 냉각재 상실사고에 대하여 REMIX 로드, 증기관 파단사고에 대하여는 COM-MIX-1B 코드에 의한 계산과 비교되었다.

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