• Title, Summary, Keyword: Reactor Pressure Vessel

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DETAILED EVALUATION OF THE IN-VESSEL SEVERE ACCIDENT MANAGEMENT STRATEGY FOR SBLOCA USING SCDAP/RELAP5

  • Park, Rae-Joon;Hong, Seong-Wan;Kim, Sang-Baik;Kim, hee-Dong
    • Nuclear Engineering and Technology
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    • v.41 no.7
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    • pp.921-928
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    • 2009
  • As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

Probabilistic Integrity Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock (가압열충격을 받는 원자로압력용기의 확률론적 건전성 해석)

  • Kim, Jong-Wook;Huh, Nam-Su;Yoo, Yeon-Sik;Kim, Tae-Wan
    • Proceedings of the KSME Conference
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    • pp.727-728
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    • 2008
  • The objective of this study is to evaluate the integrity for a reactor pressure vessel under the pressurized thermal shock by applying the probability fracture mechanics. A semi-elliptical axial crack is assumed to be in the beltline region of the reactor pressure vessel. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT. The probabilistic integrity analysis was performed using the Monte Carlo simulation.

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FATIGUE ANALYSIS OF A REACTOR PRESSURE VESSEL FOR SMART

  • Jhung, Myung-Jo
    • Nuclear Engineering and Technology
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    • v.44 no.6
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    • pp.683-688
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    • 2012
  • The structural integrity of mechanical components during several transients should be assured in the design stage. This requires a fatigue analysis including thermal and stress analyses. As an example, this study performs a fatigue analysis of the reactor pressure vessel of SMART during arbitrary transients. Using heat transfer coefficients determined based on the operating environments, a transient thermal analysis is performed and the results are applied to a finite element model along with the pressure to calculate the stresses. The total stress intensity range and cumulative fatigue usage factor are investigated to determine the adequacy of the design.

The Study on Sizing of the Pressure Relief Valve for Overpressure Protection of a Reactor Pressure Vessel in Low Temperature Condition (저온 상태의 원자로 압력용기의 과압방지를 위한 압력방출밸브 용량 결정에 관한 연구)

  • Lee, Jun;Kim, Yoo
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.4 no.2
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    • pp.7-12
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    • 2008
  • The purpose of this study is to present a methodology to estimate the capacity of the pressure relief valve which prevents overpressure of the pressure vessel in a cold state. In this methodology, the transient behavior of the flow rate through the pressure relief valve and the pressure inside the pressure vessel are considered. The result of this study shows the followings; The more the relief valve capacity is considered in excess, the more the initial relief flow rate and the initial pressure inside the pressure vessel are high and low respectively. When the relief valve capacity is determined properly, the pressure inside the pressure vessel maintains almost the same value, so the ASME code requirement will be met.

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Analysis of Chemistry Factor and RTPTS Margin for Domestic Reactor Pressure Vessel Materials by using the Surveillance Data (감시시험 결과를 이용한 국내원전 압력용기 재료의 Chemistry Factor 및 RTPTS 평가여유도 분석)

  • Lee, Ho-Jin;Yoon, Ji-Hyun;Choi, Kwon-Jae;Lee, Bong-Sang
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.7 no.3
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    • pp.15-22
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    • 2011
  • The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

Vessel failure sensitivities of an advanced reactor for SBLOCA

  • Jhung, Myung Jo;Oh, Chang-Sik;Choi, Youngin;Kang, Sung-Sik
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.185-191
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    • 2020
  • Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×1019 n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

Structural design and integrity evaluations for reactor vessel of PGSFR sodium-cooled fast reactor (PGSFR 소듐냉각고속로 원자로용기 설계 및 구조건전성 평가)

  • Koo, Gyeong Hoi;Kim, Sung Kyun
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.12 no.1
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    • pp.70-77
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    • 2016
  • In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

Neutron Fluence Evaluation for Reactor Pressure Vessel Using 3D Discrete Ordinates Transport Code RAPTOR-M3G (3차원 수송계산 코드(RAPTOR-M3G)를 이용한 원자로 압력용기 중성자 조사량 평가)

  • Maeng, Young Jae;Lim, Mi Joung;Kim, Byoung Chul
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.10 no.1
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    • pp.107-112
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    • 2014
  • The Code of Federal Regulations, Title 10, Part 50, Appendix H requires surveillance program for reactor pressure vessel(RPV) that the peak neutron fluence at the end of the design life of the vessel will exceed $1.0E+17n/cm^2$ (E>1.0MeV). 2D/1D Synthesis method based on DORT 3.1 transport calculation code has been widely used to determine fast neutron(E>1.0MeV) fluence exposure to RPV in the beltline region. RAPTOR-M3G(RApid Parallel Transport Of Radiation-Multiple 3D Geometries) performing full 3D transport calculation was developed by Westinghouse and KRIST(Korea Reactor Integrity Surveillance Technology) and applied for the evaluations of In-Vessel and Ex-Vessel neutron dosimetry. The reaction rates from measurement and calculation were compared and the results show good agreements each other.

Pressure-temperature limit curve for reactor vessel evaluated by ASME code

  • Jhung, Myung Jo;Kim, Seok Hun;Jung, Sung Gyu
    • Structural Engineering and Mechanics
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    • v.14 no.2
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    • pp.191-208
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    • 2002
  • A comparative assessment study for a generation of the pressure-temperature (P-T) limit curve of a reactor vessel is performed in accordance with ASME code. Using cooling or heating rate and vessel material properties, stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during reactor cool-down and heat-up. P-T limit curves are analyzed with respect to defect orientation, clad thickness, toughness curve, cooling or heating rate and neutron fluence. The resulting P-T curves are compared each other.

A numerical study on convective heat transfer characteristics at the vessel surface of the Korean Next Generation Reactor (차세대 원자로 용기내 vessel 내면에서의 대류 열전달특성에 관한 수치해석적 연구)

  • Jung, S.D.;Kim, C.N.
    • Proceedings of the KSME Conference
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    • pp.228-233
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    • 2000
  • The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

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