Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C17/00—Monitoring; Testing ; Maintaining
  • G21C17/02—Devices or arrangements for monitoring coolant or moderator
  • G21C17/022—Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
  • G21C17/0225—Chemical surface treatment, e.g. corrosion
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C17/00—Monitoring; Testing ; Maintaining
  • G21C17/02—Devices or arrangements for monitoring coolant or moderator
  • G21C17/022—Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
  • G21C19/28—Arrangements for introducing fluent material into the reactor core; Arrangements for removing fluent material from the reactor core
  • G21C19/30—Arrangements for introducing fluent material into the reactor core; Arrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products deterioration or corrosion products, impurities, e.g. by cold traps
  • G21C19/307—Arrangements for introducing fluent material into the reactor core; Arrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products deterioration or corrosion products, impurities, e.g. by cold traps specially adapted for liquids
• • 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

• high-temperature water means water having a temperature of about 150 ⁇ C or greater, steam, or the condensate thereof.
• High- temperature water can be found in a variety of known apparatus, such as water deaerators, nuclear reactors, and steam-driven power plants.
• a reactor pressure ves ⁇ sel contains the reactor coolant, i.e. water, which removes heat from the nuclear core.
• Respective piping circuits carry the heated water or steam to the steam generators or turbines and carry circulated water or feedwater back to the vessel.
• Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288 ⁇ C for a boiling water reactor (B R) , and about 15 MPa and 320°C for a pressurized water reactor (P R) .
• B R boiling water reactor
• P R pressurized water reactor
• Some of the materials exposed to high-temperature water include carbon steel, alloy steel, stainless steel, and nickel-based, cobalt-based and zirconium-based alloys.
• corrosion occurs on the materials exposed to the high-temperature water.
• Such corrosion contributes to a variety of problems, e.g., stress corrosion cracking, crevice corrosion, erosion corrosion, sticking of pressure relief valves and buildup of the gamma radiation-emitting Co-60 isotope.
• Stress corrosion cracking is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds, exposed to high- temperature water.
• the ECP is a measure of the thermodynamic ten ⁇ dency for corrosion phenomena to occur, and is a funda ⁇ mental parameter in determining rates of, e.g., SCC, corrosion fatigue, corrosion film thickening, and gen ⁇ eral corrosion.
• SCC corrosion fatigue
• gen ⁇ eral corrosion e.g., SCC, corrosion fatigue, corrosion film thickening, and gen ⁇ eral corrosion.
• the radiolysis of the primary water cool ⁇ ant in the reactor core causes the net decomposition of a small fraction of the water to the chemical products H 2 , H 2 0 2 , 0 2 and oxidizing and reducing radicals.
• equilibrium concen- trations of 0 2 , H 2 0 2 , and H 2 are established in both the water which is recirculated and the steam going to the turbine.
• IGSCC of Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni and 2% Mn) used in BWRs can be mitigated by reducing the ECP of the stainless steel to values below -230 mV(SHE) .
• An effective method of achieving this objective is to use HWC.
• high hydrogen additions e.g., of about 200 ppb or greater, that may be required to reduce the ECP below the critical potential, can result in a higher radiation level in the steam-driven turbine section from incorporation of the short-lived N-16 species in the steam.
• the amount of hydrogen addition required to provide mitigation of IGSCC of pressure vessel internal components results in an increase in the main steam line radiation monitor by a factor of five to eight. This increase in main steam line radiation can cause high, even unacceptable, environmental dose rates that can require expensive investments in shield ⁇ ing and radiation exposure control.
• recent investigations have focused on using minimum levels of hydrogen to achieve the benefits of HWC with minimum increase in the main steam radiation dose rates.
• FIG. 5 is a graph showing corrosion potential measurements for electrodes in 550°F water containing various amounts of oxygen, the electrodes being made from
• Each fuel assembly is supported at the top by top guide 19 and at the bottom by core plate 21. Water flowing through down- comer annulus 16 then flows to the core lower plenum 24. The water subsequently enters the fuel assemblies 22 disposed within core 20, wherein a boiling boundary layer (not shown) is established.
• a mixture of water and steam enters core upper plenum 26 under shroud head 28.
• Core upper plenum 26 provides standoff between the steam—water mixture exiting core 20 and entering verti- cal standpipes 30, which are disposed atop shroud head 28 and in fluid communication with core upper plenum 26.
• the steam-water mixture flows through standpipes 30 and enters steam separators 32, which are of the axial-flow centrifugal type.
• the BWR has two recirculation pumps, each of which provides the driving flow for a plurality of jet pump assemblies.
• the pressurized driving water is supplied to each jet pump nozzle 44 via an inlet riser 47, an elbow 48 and an inlet mixer 46 in flow sequence.
• a typical BWR has 16 to 24 inlet mixers.
• the zirconium compound is injected at a point upstream of the feedwater inlet 12 (see FIG. 1) .
• ECP and 0 2 test data at 547°F for a Type 304 stainless steel CERT specimen held in place in a clevis using oxidized Z-Nb pins, a Type 304 stainless steel electrode tip and a Type 304 stainless steel CERT specimen held in a clevis using Zr0 2 (MgO) ceramic pins are com ⁇ pared in Table I. All stainless steel specimens had been pre-oxidized before the test.
• a constant extension rate tensile (CERT) test was performed at 547 ⁇ F with a Type 304 stainless steel specimen.
• the specimen was held in the clevis of a standard CERT autoclave using oxidized Zr—Nb pins.
• the ECP of the stainless steel specimen was far more negative (-196 mV/SHE) than expected at the oxygen level (225 ppb 0 2 ) used in the study.
• a preoxidized Type 304 stainless steel electrode tip that was in the same autoclave showed a potential of +60 mV(SHE) , which was antici ⁇ pated in the high-oxygen environment used.
• Zirconium doping of stainless steel (or other metal) components of a BWR by injecting a zirconium compound into the high-temperature water would make it possible to polarize the stainless steel potential in the negative direction without using hydrogen.
• the benefits of this achievement would be numerous.
• the main steam radiation dose rates should remain at the background level because no hydrogen will be used.
• zirconium and its alloys are compatible with fuel cladding material and hence fuel removal may not be required during zirconium doping.
• the cost of zirconium is much less than the cost of palladium.
• zirconium doping can be performed in situ either during shutdown (when the water temperature inside the reactor is about 40- 60 ⁇ C) or during operation (when the water temperature inside the reactor is about 288°C).
• all structural surfaces exposed to the recir ⁇ culating water carrying the injected zirconium will be doped.
• structural reactor components can be treated ex situ before installation in the reactor. An experiment was performed to test the effect on corrosion potential of exposing Type 304 stainless steel to a Zr0(N0 3 ) 2 solution. Test specimens of Type 304 stainless steel (W" diam.
• FIG. 3 shows an Auger electron spectroscopy depth profile of the surface of Type 304 stainless steel after exposure to a 1 m ZrO(N0 3 ) 2/ solution at 60°C for 10 days.
• the data in FIG. 3 confirm that zirconium has been incorporated into the oxide film as a result of the treatment in accordance with the invention. Zirconium was incorporated into the oxide film to a depth of 300 A
• the Z-doped Type 304 stainless steel test specimens showed lower corrosion potentials than the undoped specimens at the same oxygen level. This difference in the corrosion potentials of the Z-doped and undoped stainless steel electrodes is attributable to the change in the electrical conductivity of the oxide film caused by doping of zirconium into the oxide. By contrast, the corrosion potential of pure zirconium was about -650 V(SHE), even at high oxygen levels. As seen in FIG. 4, the corrosion potential of Z-doped Type 304 stainless steel is further reduced as the duration of the doping treatment is increased from 10 days to 20 days.
• An exemplary zirconium acetylacetonate injection solution was prepared by dissolving 52.6 mg of zirconium acetylacetonate powder in 40 ml of ethanol. The ethanol solution was then diluted with water. After dilution, 10 ml of ethanol are added to the solution. This sol ⁇ ution is then diluted with water to a volume of 1 liter. Obviously, the concentration range can be varied. Alternatively, a water-based suspension can be formed, without using ethanol, by mixing zirconium acetylace ⁇ tonate powder in water.
• the zirconium acetylacetonate compound dissolved in the ethanol/water mixture, was injected into the inlet side of the main pump in the flow loop using an injection pump at a rate so that the solution entering the autoclave (at 550 ⁇ F) had a Zr concentration of -100 ppb.
• the results of this experiment are depicted in FIG. 5.
• the corrosion potential of the Type 304 stainless steel specimen was tested in high-temperature water at 550°F.
• the response of the Zr-doped specimen was tested at different oxygen levels.
• FIG. 5 shows that the ECP of the Zr-doped specimen was negative from the outset even in the presence of high oxygen levels.
• the ECP of an undoped Type 304 stainless steel specimen pre-oxidized in 8 ppm 0 2 for one week at 550°F drops to a value of only -39 mV(SHE) even when the H 2 /0 2 molar ratio is increased to 8.5.
• the Type 304 stainless steel specimen doped with zirconium acetylacetonate shows a negative potential of -87 mV(SHE) at a dissolved oxygen concentration of 338 ppb without any hydrogen.
• zirconium doping of the stainless steel surface is extremely beneficial in reducing the ECP of the specimen and hence in controlling crack initiation and growth in stainless steel, since ECP is a primary factor that controls SCC of stainless steel and other alloys used in a nuclear reactor.
• the non-noble metals identified above as being useful in the invention can be used alone or in combination.
• the doping technique of the invention is not restricted to use with stainless steel surfaces, but also has application in reducing the ECP of other metals which are susceptible to IGSCC, e.g., nickel- based alloys, carbon steel and low alloy steels.
• An alternative application technology includes having the metal compound as pressed pellets in a basket hung inside the reactor at different locations and operating the reactor with pump heat until metal doping occurs. Another approach would be to inject the compound locally into areas that have a higher susceptibility to IGSCC. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter.

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• Physics & Mathematics (AREA)
• Engineering & Computer Science (AREA)
• Plasma & Fusion (AREA)
• General Engineering & Computer Science (AREA)
• High Energy & Nuclear Physics (AREA)
• Preventing Corrosion Or Incrustation Of Metals (AREA)
• Inorganic Compounds Of Heavy Metals (AREA)
• Chemical Treatment Of Metals (AREA)

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• 1995
  • 1995-06-23 EP EP95928071A patent/EP0715762A1/de not_active Withdrawn
  • 1995-06-23 WO PCT/US1995/008905 patent/WO1996000447A1/en not_active Application Discontinuation
  • 1995-06-23 JP JP50350296A patent/JP3749731B2/ja not_active Expired - Fee Related
  • 1995-06-23 KR KR1019960700891A patent/KR100380127B1/ko not_active IP Right Cessation

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