US20050135545A1 - Thermal load reducing system for nuclear reactor vessel - Google Patents

Thermal load reducing system for nuclear reactor vessel Download PDF

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Publication number
US20050135545A1
US20050135545A1 US10/682,859 US68285903A US2005135545A1 US 20050135545 A1 US20050135545 A1 US 20050135545A1 US 68285903 A US68285903 A US 68285903A US 2005135545 A1 US2005135545 A1 US 2005135545A1
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United States
Prior art keywords
reactor vessel
heat conductive
reactor
wall
thermal load
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Abandoned
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US10/682,859
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English (en)
Inventor
Naoto Kasahara
Masanori Ando
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Japan Atomic Energy Agency
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Japan Nuclear Cycle Development Institute
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Filing date
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Application filed by Japan Nuclear Cycle Development Institute filed Critical Japan Nuclear Cycle Development Institute
Assigned to JAPAN NUCLEAR CYCLE DEVELOPMENT INSTITUTE reassignment JAPAN NUCLEAR CYCLE DEVELOPMENT INSTITUTE ASSIGNMENT OF ASSIGNORS INTEREST (SEE DOCUMENT FOR DETAILS). Assignors: ANDO, MASANORI, KASAHARA, NAOTO
Publication of US20050135545A1 publication Critical patent/US20050135545A1/en
Priority to US11/785,126 priority Critical patent/US20070280399A1/en
Priority to US11/806,723 priority patent/US8036335B2/en
Abandoned legal-status Critical Current

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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C11/00—Shielding structurally associated with the reactor
    • G21C11/08—Thermal shields; Thermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield ; Thermal insulation
    • G21C11/083—Thermal shields; Thermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield ; Thermal insulation consisting of one or more metallic layers
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C13/00—Pressure vessels; Containment vessels; Containment in general
    • G21C13/02—Details
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors

Definitions

  • the present invention relates to a thermal load reducing system for a nuclear reactor vessel, available for the reduction of thermal load near a coolant liquid surface of a reactor vessel and for the reduction of thermal load near a temperature stratified layer in a reactor vessel.
  • a reactor vessel in a fast breeder reactor is supported at its upper end by a concrete wall, which must be maintained at the temperature of 100° C. or lower. Because it has a high-temperature coolant at 550° C. or higher in a plenum above reactor core, there occurs a steep temperature gradient in the vertical direction from the coolant liquid surface to the upper supported end. In particular, during starting operation, both temperature and liquid level rise at the same time, the gradient becomes steeper. As a result, high thermal stress develops, in principle, on the reactor wall near the liquid surface where the temperature gradient deflects.
  • the conventional method for reducing thermal load has its principal aims to prevent the rise of liquid level using a liquid level controller, to evenly cool down the reactor wall using a reactor wall cooling system, and to decrease bending stress by designing in a thin-wall structure.
  • the liquid level controller and the reactor wall cooling system result in higher cost because of the increase of system components.
  • For designing the system in a thin-wall structure there was a limitation due to the possibility of other failure modes.
  • the method described in the patent document referred to above also leads to higher cost due to the increase of system components.
  • the object of the present invention is to provide sure operations and to contribute, without giving a significant impact to the construction cost, both for the increased safety of the reactors and for the improved economy of the plant by reducing the thermal load itself, being the cause to generate stress.
  • the present invention is characterized in that a heat conductive member is installed outside the reactor vessel in the area above and below the coolant liquid surface not contacting to the reactor vessel wall.
  • the present invention is characterized in that aforementioned heat conductive member is a plate supported by a guard vessel.
  • the present invention is characterized in that the aforementioned heat conductive member is the guard vessel wall.
  • the present invention is characterized in that the aforementioned heat conductive member is of better material in heat conductivity than that of the reactor vessel.
  • the present invention is characterized in that the aforementioned good heat conductive material is high chrome steel.
  • the present invention is characterized in that the aforementioned good heat conductive material is graphite.
  • FIG. 1 is a drawing to show an embodiment example of a thermal load reducing system for reduction of stress near the liquid surface of a reactor vessel;
  • FIG. 2 is a drawing to explain the principle of thermal load reduction near a coolant liquid surface of a reactor vessel
  • FIG. 3 is a drawing to show the relationship between equivalent heat transfer coefficient due to radiation and temperature in use
  • FIG. 4 is a drawing to show an analysis mesh model
  • FIG. 5 is a drawing to show an analysis mesh model zoomed in at the part of heat conductive plate
  • FIG. 6 is a drawing to show the difference in generated stresses obtained as the result of the analysis between the cases with or without heat conductive plate;
  • FIG. 7 is a drawing to show another embodiment example of the thermal load reducing system for reduction of stress near the liquid surface of the reactor vessel;
  • FIG. 8 is a drawing to explain the principle of thermal load reduction near the coolant liquid surface of the reactor vessel
  • FIG. 9 is a drawing to show the relationship between equivalent heat transfer coefficient due to radiation and temperature in use.
  • FIG. 10 is a drawing to show an analysis mesh model
  • FIG. 11 is a drawing to show an analysis mesh model zoomed in at the part near the liquid surface.
  • FIG. 12 is a drawing to show the difference in generated stresses obtained as the result of the analysis.
  • FIG. 1 shows an embodiment example of a thermal load reducing system for reduction of stress near the liquid surface of a reactor vessel.
  • a guard vessel 2 is provided outside the wall of the reactor vessel 1 to catch the coolant in the unlikely event of coolant leakage.
  • inert gas is filled for the protection of the reactor vessel.
  • heat insulating material 8 is provided on the outer wall of the guard vessel to keep the concrete temperature not going up.
  • the reactor vessel inner wall above the coolant surface 9 covered with heat insulating material 10 , is heat insulated from high temperature of the coolant and this part of the reactor wall is kept at low temperature. Therefore, between high temperature reactor wall below the coolant surface and low temperature reactor wall above the coolant surface, a temperature distribution in the vertical direction takes place, causing a thermal stress.
  • a heat conductive member comprised of a good heat conductive material such as graphite etc., stable for a long time, is positioned.
  • the present embodiment is characterized in that the heat conductive plate 20 as the heat conductive member, is positioned without contact, to expedite the thermal conduction in the vertical direction of the reactor vessel wall and to reduce the thermal stress.
  • the heat conductive plate 20 as shown in the drawing is supported by structures outside reactor vessel such as the guard vessel.
  • the dimensions of the heat conductive plate should approximately be 30 mm in thickness, 1500 mm in length, and positioned at the distance of about 60 mm measured from the reactor vessel wall to the surface of the plate, vertically positioned longer below the liquid level than above the liquid level, for example about 500 mm long above the liquid level and about 1000 mm long below the liquid level.
  • FIG. 2 is a drawing to explain the principle of thermal load reduction near the coolant liquid surface of the reactor vessel; FIG. 2 ( a ) for the case without the heat conductive plate and FIG. 2 ( b ) for the case with the heat conductive plate.
  • the reactor vessel in a fast breeder reactor is supported by concrete structure, the upper end of which must be maintained at the temperature of 100° C. or lower.
  • the temperature of the contained coolant rises from 200° C. to 550° C.
  • a local temperature gradient in the vertical direction developed during this process generates a high thermal stress on the reactor wall. Specifically, in case temperature distribution during the starting of the reactor vessel is left freely as it goes ( FIG.
  • FIG. 3 is a drawing to show the relationship between equivalent heat transfer coefficient due to radiation and temperature in use.
  • the thermal radiation/conduction quantity between the parallel planes is expressed in the following Formula 1.
  • q ⁇ eq ⁇ ⁇ ⁇ ( T 1 4 - T 2 4 )
  • ⁇ ⁇ ⁇ eq ⁇ ⁇ ⁇ ( T 1 3 + T 1 2 ⁇ T 2 + T 1 ⁇ T 2 2 + T 2 3 ) ⁇ ( T 1 - T 2 )
  • ⁇ 1 and ⁇ 2 is emission rate dependent on materials (quantity ratio in heat received versus heat radiated), between reactor wall made of stainless steal and high chrome steel e.g.
  • Table 2 shows physical values of a heat conductive plate made of graphite
  • Table 3 shows the same but made of 12% Cr steel, respectively.
  • the analysis was carried out with the internal sodium temperature raised from 200° C. up to 600° C.
  • the heating up speed was at a rate of 15° C./hr from 200 C up to 400 C, and 20° C./hr from 400° C. up to 600° C.
  • the rise in liquid level to the temperature rise of sodium had also been taken into consideration.
  • the rise in liquid sodium as against the rise in temperature were assumed for 880 mm for the temperature range from 200° C. up to 400° C., and thereafter for 350 mm from 400° C. up to 600° C., respectively.
  • a mesh generation program for finite element analysis Femap v7.1 and for analytic tool a general purpose nonlinear structural analysis system FINAS v14 had been utilized.
  • FIG. 4 shows the analysis mesh model made out of FIG. 1
  • FIG. 5 shows the same zoomed in on the heat conductive plate, respectively.
  • Table 5 shows the list of elements used for the analysis. TABLE 5 Elements used for developing finite element model Heat Heat conductive 8 node tetragon axisymmetric element Transfer (HQAX8) Analysis Heat conductive 3 node axisymmetric element (FCAX3) Radiation link 6 node tetragon axisymmetric element (RALINK6) Stress 8 node tetragon axisymmetric element (QAX8) Analysis Elements Used for Developing Finite Element Model
  • FCAX3 Heat conductive 3 node axisymmetric element
  • FIG. 6 is a drawing to show the difference in generated stresses obtained as the result of the analysis among the 3 patterns of analysis; a case without thermal stress reduction by means of a heat conductive plate, a case with graphite heat conductive plate and a case with a 12% Cr steel heat conductive plate.
  • horizontal axis indicates generated stress Sn (MPa) and vertical axis indicates vertical coordinate (mm).
  • the stress intensity range (Sn) on the reactor outer wall being utilized as the strength designing parameter, had been calculated and indicated.
  • the calculation result proves that Sn varies to the difference between emission rate and heat transfer coefficient of the heat conductive plate.
  • the result of an analysis using graphite which is good in emission, as heat conductive plate showed the maximum value of Sn reduced from approx 590 MPa to approx 430 MPa, or by about 27%.
  • the result using 12% Cr steel as heat conductive plate the maximum value of Sn was confirmed to have reduced to approx 500 MPa or by about 15%. This verified that a simple installation using a heat conductive plate, significantly reduced the thermal stress taking place on the reactor wall.
  • FIG. 7 shows another embodiment example of the thermal load reducing system for reduction of stress near the liquid level of a reactor vessel.
  • guard vessel is made of the same material with the reactor vessel.
  • the material of the guard vessel 2 being altered to better material in heat conduction coefficient than that of the reactor vessel.
  • This embodiment is characterized in that the guard vessel 2 become a heat conductive member, and the guard vessel thermally combined by radiation with the reactor wall, expediting the thermal conduction in the vertical direction to the reactor vessel wall, reduces the thermal stress near the liquid level.
  • the composition of the reactor vessel is identical to that of FIG. 1 except that there is no heat conductive plate.
  • FIG. 8 is a drawing to explain the principle of thermal load reduction near the coolant liquid surface of a reactor vessel; FIG. 8 ( a ) for the case without a guard vessel and FIG. 8 ( b ) for the case with guard vessel of good heat conductive material.
  • the guard vessel of good heat conductive material is heated up with the heat radiation from the high temperature reactor wall below the liquid surface, and in turn, the low temperature reactor wall above the liquid surface is heated up with the heat radiation from the guard vessel of good heat conductive material.
  • This makes it possible to reduce the temperature gradient in the vertical direction, which causes the stress.
  • the temperature gradient at point S where the maximum stress taking place is smoothed at the time T when the maximum stress being generated, and the thermal load is reduced.
  • This embodiment is characterized in that the method not only avoids any effect to the construction cost with no new members being added, but also assures operation due to its non-contacting and static structure.
  • FIG. 9 is a drawing to show the relationship between equivalent heat transfer coefficient due to radiation and temperature in use, horizontal axis indicates temperature (T) of reactor wall and guard vessel (of high chrome steel such as 12% Cr steel), while vertical axis indicates equivalent heat transfer coefficient (heq).
  • FIG. 10 shows an analysis mesh model made out of FIG. 7
  • FIG. 11 shows an analysis mesh model zoomed-in at the liquid surface part.
  • FIG. 12 (corresponding to FIG. 6 ) is a drawing to show the difference in generated stresses obtained as the result of the analysis.
  • Horizontal axis indicates generated stress Sn (MPa) and vertical axis indicates vertical coordinate (mm), showing the analysis results in 3 patterns of without considering the radiation heating by guard vessel, with a guard vessel of 12% Cr Steel, and with a guard vessel of 316FR stainless steel.
  • the analysis results were indicated, in accordance with vertical distribution of the stress intensity area (Sn) along the reactor outer wall, which is utilized as the strength evaluation parameter in respective design.
  • the maximum value of Sn is reduced from approx 590 MPa to approx 519 MPa, or by about 12%. Also, the analysis result using 316 FR stainless steel as guard vessel, the maximum value of Sn was confirmed to have reduced to approx 562 MPa or by about 4.7%. This verified that a change in guard vessel material from ordinary one to good heat conductive one, achieves a distinguished thermal load reduction.
  • thermal load near the coolant liquid level can be reduced by placing a heat conductive member, outside the reactor vessel in the area above and below the coolant surface without contacting to the reactor vessel wall.
  • thermal load near the coolant surface can be reduced, not only at a low cost but also assuring the operation since it is a non-contacting and static structure.
  • thermal load near the coolant surface can be reduced, without adding any member anew, not affecting the construction cost and assuring the operation since it is a non-contacting and static structure.

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  • Engineering & Computer Science (AREA)
  • Physics & Mathematics (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Life Sciences & Earth Sciences (AREA)
  • Health & Medical Sciences (AREA)
  • Biomedical Technology (AREA)
  • General Health & Medical Sciences (AREA)
  • Molecular Biology (AREA)
  • Monitoring And Testing Of Nuclear Reactors (AREA)
  • Physical Or Chemical Processes And Apparatus (AREA)
  • Furnace Housings, Linings, Walls, And Ceilings (AREA)
US10/682,859 2003-03-04 2003-10-14 Thermal load reducing system for nuclear reactor vessel Abandoned US20050135545A1 (en)

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US11/785,126 US20070280399A1 (en) 2003-03-04 2007-04-16 Thermal load reducing system for nuclear reactor vessel
US11/806,723 US8036335B2 (en) 2003-03-04 2007-06-04 Thermal load reducing system for nuclear reactor vessel

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JP2003057147A JP3909700B2 (ja) 2003-03-04 2003-03-04 原子炉容器の熱荷重緩和装置
JP2003-057147 2003-03-04

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CN102260520A (zh) * 2010-05-31 2011-11-30 华东理工大学 一种新型急冷废热锅炉下管箱及其内衬刚玉的设计方法

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US3186913A (en) * 1960-05-18 1965-06-01 North American Aviation Inc Graphite moderated nuclear reactor
US3650709A (en) * 1965-06-22 1972-03-21 Avesta Jernverks Ab Ferritic, austenitic, martensitic stainless steel
US3769161A (en) * 1970-07-09 1973-10-30 Commissariat Energie Atomique Reactor vessel supporting device
US3764835A (en) * 1972-05-25 1973-10-09 Massachusetts Inst Technology Double shielded superconducting field winding
US4162229A (en) * 1976-04-07 1979-07-24 Gesellschaft zur Forderung der Forschung an der Eidgenosslschen Technischen Hochschule Decontamination process
US4172011A (en) * 1976-08-12 1979-10-23 Nuclear Power Company Limited Liquid metal cooled nuclear reactor constructions
US4302296A (en) * 1978-09-28 1981-11-24 Electric Power Research Institute, Inc. Apparatus for insulating hot sodium in pool-type nuclear reactors
US4362694A (en) * 1979-07-17 1982-12-07 Commissariat A L'energie Atomique Liquid metal-cooled nuclear reactor
US4696791A (en) * 1984-07-17 1987-09-29 Sulzer Brothers Limited Nuclear reactor installation
US4780270A (en) * 1986-08-13 1988-10-25 The United States Of America As Represented By The United States Department Of Energy Passive shut-down heat removal system
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US4859402A (en) * 1987-09-10 1989-08-22 Westinghouse Electric Corp. Bottom supported liquid metal nuclear reactor
US5229067A (en) * 1989-11-17 1993-07-20 Siemens Aktiengesellschaft Liquid metal cooled nuclear reactor
US5583900A (en) * 1993-03-18 1996-12-10 Hitachi, Ltd. Structural member having superior resistance to neutron irradiation embrittlement, austenitic steel for use in same, and use thereof
US5774514A (en) * 1993-10-29 1998-06-30 Rubbia; Carlo Energy amplifier for nuclear energy production driven by a particle beam accelerator
US5406602A (en) * 1994-04-15 1995-04-11 General Electric Company Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability
US5699394A (en) * 1995-07-13 1997-12-16 Westinghouse Electric Corporation Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

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Publication number Publication date
JP3909700B2 (ja) 2007-04-25
US20080107226A1 (en) 2008-05-08
JP2004264241A (ja) 2004-09-24
US8036335B2 (en) 2011-10-11
FR2852139B1 (fr) 2005-05-27
FR2852139A1 (fr) 2004-09-10
US20070280399A1 (en) 2007-12-06

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