• Title, Summary, Keyword: Reactor Vessel

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Integrity of the Reactor Vessel Support System for a Postulated Reactor Vessel Closure Head Drop Event

  • Kim, Tae-Wan;Lee, Ki-Young;Lee, Dae-Hee;Kim, Kang-Soo
    • Nuclear Engineering and Technology
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    • v.28 no.6
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    • pp.576-582
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    • 1996
  • The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

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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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Computational Study of the Mixed Cooling Effects on the In-Vessel Retention of a Molten Pool in a Nuclear Reactor

  • Kim, Byung-Seok;Ahn, Kwang-Il;Sohn, Chang-Hyun
    • Journal of Mechanical Science and Technology
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    • v.18 no.6
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    • pp.990-1001
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    • 2004
  • The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

CORIUM BEHAVIOR IN THE LOWER PLENUM OF THE REACTOR VESSEL UNDER IVR-ERVC CONDITION: TECHNICAL ISSUES

  • Park, Rae-Joon;Kang, Kyoung-Ho;Hong, Seong-Wan;Kim, Sang-Baik;Song, Jin-Ho
    • Nuclear Engineering and Technology
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    • v.44 no.3
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    • pp.237-248
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    • 2012
  • Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

An Investigation of Thermal Margin for External Reactor Vessel Cooling(ERVC) in Large Advanced Light Water Reactors(ALWR)

  • Park, Jong-Woon;Jerng, Dong-Wook
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.473-478
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    • 1997
  • A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.

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Reactor Vessel Water Level Estimation During Severe Accidents Using Cascaded Fuzzy Neural Networks

  • Kim, Dong Yeong;Yoo, Kwae Hwan;Choi, Geon Pil;Back, Ju Hyun;Na, Man Gyun
    • Nuclear Engineering and Technology
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    • v.48 no.3
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    • pp.702-710
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    • 2016
  • Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

PREDICTION OF THE REACTOR VESSEL WATER LEVEL USING FUZZY NEURAL NETWORKS IN SEVERE ACCIDENT CIRCUMSTANCES OF NPPS

  • Park, Soon Ho;Kim, Dae Seop;Kim, Jae Hwan;Na, Man Gyun
    • Nuclear Engineering and Technology
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    • v.46 no.3
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    • pp.373-380
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    • 2014
  • Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

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.

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.

Hydraulic and Structural Analysis for APR1400 Reactor Vessel Internals against Hydraulic Load Induced by Turbulence

  • Kim, Kyu Hyung;Ko, Do Young;Kim, Tae Soon
    • International Journal of Safety
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    • v.10 no.2
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    • pp.1-5
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    • 2011
  • The structural integrity assessment of APR1400 (Advanced Power Reactor 1400) reactor vessel internals has been being performed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program prior to commercial operation. The program is composed of a hydraulic and structural analysis, a vibration measurement, and an inspection. This paper describes the hydraulic and structural analysis on the reactor vessel internals due to hydraulic loads caused by the turbulence of reactor coolant. Three-dimensional models were built for the hydraulic and structural analysis and then hydraulic loads and structural responses were predicted for five analysis cases with CFX and ANSYS respectively. The structural responses show that the APR1400 reactor vessel internals have sufficient structural integrity in comparison with the acceptance criteria.

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