JP2007017245A - Method for mitigating stress corrosion cracking in structural material for nuclear power plant - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To improve integrity of a plant by suppressing increase in stress corrosion cracking sensitivity by suppressing increase in electric conductivity. <P>SOLUTION: In an operation of a boiling water nuclear power plant, a material for reducing the concentration of oxygen or hydrogen peroxide in reactor water by chemically reacting with the oxygen or hydrogen peroxide is injected into the reactor water. During stop of the operation, a layer of a compound having a spinel type structure composed of bivalent metal ions and trivalent iron ions is formed on a heat transfer surface of a feed water heater. Even when hydrogen injection or the like to the reactor water is stopped, increase in chrome ion concentration in the reactor water can be suppressed. As a result, the electric conductivity can be suppressed, and increase in sensitivity to stress corrosion cracking can be suppressed. <P>COPYRIGHT: (C)2007,JPO&INPIT

Description

  More particularly, the present invention relates to a method for mitigating stress corrosion cracking of nuclear reactor structural materials such as stainless steel and nickel-base alloys.

  In a boiling water nuclear power plant, it is an important issue to suppress stress corrosion cracking (hereinafter referred to as “SCC”) with respect to the materials constituting the reactor internal structure and the pressure boundary from the viewpoint of improving the plant operating rate. Yes. SCC is supposed to occur when three factors of material, stress and environment overlap. Therefore, SCC can be mitigated by mitigating at least one of these three factors. Among the three factors, mitigation of environmental factors has the advantage that it can generally be implemented with less construction and lower costs.

  During plant operation, the reactor cooling water undergoes radiolysis due to the action of strong gamma rays and neutron rays in the core, producing oxygen and hydrogen peroxide. As a result, the reactor cooling water has a high temperature environment in which oxygen and hydrogen peroxide are present in the order of several hundreds of ppb (100 ° C. or higher is called high temperature. Note that the temperature at the core outlet during rated power operation is 288 ° C.). . The structural material (stainless steel, nickel-base alloy, etc.) constituting the reactor internal structure and the pressure boundary is exposed to the reactor cooling water in such an environment.

  There is a corrosion potential (hereinafter referred to as “ECP”) as an index representing the environment of the reactor cooling water. FIG. 2 shows the relationship between the crack growth rate (hereinafter referred to as “CGR”) of 304 type stainless steel (hereinafter referred to as “SUS304 steel”) and ECP. The horizontal axis in FIG. 2 indicates ECP, and the vertical axis indicates CGR. SUS304 steel was used as a test piece, and the test was performed under the conditions of a temperature of 288 ° C. and an electric conductivity of 0.1 μS / cm to 0.3 μS / cm. FIG. 2 shows that CGR decreases as ECP decreases.

  FIG. 3 shows the dependence of ECP for SUS304 steel on oxygen or hydrogen peroxide concentration. The horizontal axis in FIG. 3 indicates the concentration of oxygen or hydrogen peroxide, and the vertical axis indicates ECP. SUS304 steel was used as a test piece, and the test was performed under the condition of a temperature of 280 ° C. FIG. 3 shows that the ECP decreases as the concentration of oxygen and hydrogen peroxide decreases. Therefore, it is necessary to reduce the ECP in order to mitigate the SCC of the structural material exposed to the reactor cooling water, and the reduction of the ECP includes the reactor water (including the coolant that has flowed through the reactor core). Cooling water from which impurities have not been removed by ion exchange resin tower, etc. Reactor upstream of reactor pressure vessel, reactor cooling water recirculation system piping, bottom drain piping, reactor cooling water filtration demineralizer This is achieved by reducing the concentration of oxygen and hydrogen peroxide present in the cooling water purification system piping.

  In response to this problem, a technology for adding hydrogen from the water supply system (hereinafter referred to as “hydrogen injection”) has been implemented in many plants in Japan and overseas. Hydrogen injection is a technique for reducing the concentration of oxygen and hydrogen peroxide in reactor water by reacting oxygen and hydrogen peroxide generated by radiolysis of water with water to return them to water (for example, Patent Document 1). reference).

FIG. 2 also shows the influence of electrical conductivity in addition to the relationship between CGR and ECP. FIG. 2 shows that the CGR is higher when the electrical conductivity is 0.3 μS / cm than when the electrical conductivity is 0.1 μS / cm even with the same ECP. In other words, CGR increases as the electrical conductivity increases. In plants where hydrogen is injected into the reactor water of a boiling water nuclear power plant, when hydrogen injection is stopped, the chromium ion concentration in the reactor water increases rapidly, and as a result, the electrical conductivity in the reactor water increases. Problem arises. Therefore, in a plant injecting hydrogen into reactor water, when hydrogen injection is stopped, the electrical conductivity increases due to an increase in chromium ion concentration, and as a result, SCC sensitivity may increase.

Japanese Patent Laid-Open No. 05-100087

  The present invention provides a method for mitigating stress corrosion cracking in nuclear plant structural materials that can suppress increase in stress corrosion cracking susceptibility by suppressing increase in electrical conductivity and improve plant soundness. This is the issue.

  During operation of the boiling water nuclear plant, a substance that chemically reacts with oxygen or hydrogen peroxide to reduce the concentration of oxygen or hydrogen peroxide in the reactor water is injected into the reactor water of the boiling water nuclear plant, During the shutdown period of the boiling water nuclear plant, a spinel structure compound layer composed of divalent metal ions and trivalent iron ions is formed on the heat transfer surface of the feed water heater of the boiling water nuclear plant.

  According to the present invention, an increase in the chromium ion concentration in the reactor water can be suppressed even when hydrogen injection into the reactor water is stopped. As a result, it is possible to suppress an increase in electrical conductivity, and thus it is possible to suppress an increase in sensitivity to stress corrosion cracking, thereby improving plant soundness.

  The inventor examined the method of suppressing the increase of the chromium ion concentration and thus reducing the CGR when the injection of hydrogen into the reactor water was stopped.

  In a boiling water nuclear power plant that injects hydrogen into reactor water, the phenomenon that the concentration of chromium ions in reactor water increases when hydrogen injection is stopped due to a plant shutdown operation, etc., depends on the properties of chromium ions described below. it is conceivable that.

According to Equation 1, the chromium ions exist as chromate ions (HCrO ₄ ⁻ ) in an oxidizing environment, and exist as trivalent chromium ions (Cr ³⁺ ) in a reducing environment.

(Formula 1)
HCrO ₄ ⁻ + 7H ⁺ + 3e ⁻ = Cr ³⁺ + 4H ² O

In pure water, the solubility of chromate ions is large and tends to exist as ions, but the trivalent chromium ions have low solubility and are easily precipitated as chromium oxides (Cr (OH) ₃ or Cr ₂ O ₃ ) according to Equation 2. .

(Formula 2)
Cr ³⁺ + 3H ₂ O = Cr (OH) ₃ + 3H ⁺

  Considering the properties of these chromium ions, the mechanism of the phenomenon that the chromium ion concentration in the reactor water increases when hydrogen injection is stopped in a boiling water nuclear power plant that injects hydrogen into the reactor water is as follows. Conceivable. Chromium ions are generated by corrosion elution of the heat transfer surface of the stainless steel feed water heater, and are brought into the reactor pressure vessel by cooling water (feed water). Oxygen is present in the water supply in an amount of about 10 to 100 ppb and is oxidizing, so chromium ions are present as highly soluble chromate ions. When hydrogen injection is performed, the concentration of oxygen and hydrogen peroxide in the reactor water decreases, and the reactor water becomes reducible. In reducing properties, chromium ions become trivalent chromium ions with low solubility, and thus are easily deposited as chromium oxides. That is, chromium ions are deposited and accumulated as chromium oxide on the surface of the in-furnace structural material. When hydrogen injection is stopped, the oxygen and hydrogen peroxide concentrations in the reactor water increase and the reactor water becomes oxidizing. In an oxidizing environment, chromium oxide is redissolved as highly soluble chromate ions. Since this re-dissolution occurs immediately after the hydrogen injection is stopped, it is considered that the chromium ion concentration in the reactor water increases rapidly.

  In addition, since the reactor water is oxidizing in a plant that does not perform hydrogen injection, it is considered that chromium ions exist as chromate ions and remain in the reactor water without being deposited on the surface of the structural material in the reactor. Chromium ions remaining in the reactor water are removed by a reactor water purification system (a system that draws a part of the reactor water in the reactor pressure vessel and removes impurities contained in the reactor water). Therefore, although chromium ions are constantly present in the reactor water of a plant that does not perform hydrogen injection, a rapid increase in the chromium ion concentration does not occur.

  The inventor considered the above-described mechanism regarding chromium ions, and as a result of examining a method for suppressing an increase in the chromium ion concentration when hydrogen injection into the reactor water was stopped, and thus reducing CGR, the following 4 were described. It was concluded that the kind of method was effective.

The first method is a spinel structure composed of divalent metal ions and trivalent iron ions (oxidation represented by the chemical formula XY ₂ O ₄₎ on the heat transfer surface of the feed water heater in contact with the cooling water supplied to the reactor pressure vessel. A type of crystal structure found in a product, which belongs to a cubic system and forms a compound layer of 8 chemical units (XY ₂ O ₄ ) in a unit cell.

Since the feed water heater heat transfer surface is generally 230 ° C. or lower, iron oxide (α-Fe ₂ O ₃ ) and iron hydroxide (α-FeOOH) are generated on the feed water heater heat transfer surface. Since these oxides are fine particles or amorphous, they become porous oxide layers. Therefore, the corrosion inhibitory effect with respect to the preform | base_material of a feed water heater heat transfer surface is small, and the elution of chromium ion accompanying corrosion cannot be suppressed. On the other hand, a compound having a spinel structure composed of a divalent metal ion and a trivalent iron ion is a crystalline compound and thus becomes a dense oxide layer. Therefore, if a spinel-type compound layer composed of divalent metal ions and trivalent iron ions is formed on the base material surface of the feed water heater heat transfer surface, corrosion of the feed water heater heat transfer surface base material will occur. It is possible to suppress the elution of chromium ions accompanying corrosion. In other words, by suppressing the elution of chromium ions from the heat transfer surface of the feed water heater and reducing the inflow of chromium ions into the reactor pressure vessel, the accumulation of chromium oxide on the surface of the reactor internal structure is suppressed. It is possible to suppress a rapid increase in chromium ion concentration and an increase in electrical conductivity when hydrogen injection is stopped.

It was confirmed by experiment that the corrosion of the base material of the heat transfer surface of the feed water heater can be suppressed by using a compound having a spinel structure composed of divalent metal ions and trivalent iron ions. It is shown in 1. FIG. 1 shows the results of a corrosion test of SUS304 steel on which a Fe ₃ O ₄ (compound having a spinel structure composed of divalent Fe ions and trivalent Fe ions) film was formed. Fe ₃ O ₄ layer formed was SUS304 steel sheet and Fe ₃ O ₄ of not forming a layer SUS304 steel plate (thickness of about 1 [mu] m), the temperature 260 ° C., a quasi-static environment of dissolved oxygen concentration 10 [mu] g / cm ² The corrosion test was conducted by immersing for 500 hours. The formation of the Fe ₃ O ₄ layer on the SUS304 steel plate was performed by immersing the SUS304 steel plate in a Fe ion-containing solution at a temperature of 90 ° C. in which the pH and the oxidation-reduction potential were adjusted. The vertical axis in FIG. 1 indicates the amount of weight change, and the value increases as the amount of corrosion increases. From the experimental results of FIG. 1, the corrosion amount (weight change amount) of the SUS304 steel sheet formed with Fe ₃ O ₄ was about ½ that of the untreated SUS304 steel sheet. That is, the corrosion amount was reduced to about ½ by the formation of the Fe ₃ O ₄ film. From the above experimental results, it is understood that corrosion of SUS304 steel can be suppressed by forming a spinel type compound layer on the surface of SUS304 steel.

  Here, as the divalent metal ions constituting the spinel structure compound, iron, nickel, zinc, magnesium, and manganese, which are less affected by activation and are not a source of chromium ions, are desirable.

  In the second method, zirconium or a zirconium alloy is installed on the surface of the pipe connecting the feed water heater and the pressure vessel that contacts the cooling water.

  Zirconium in zirconium or a zirconium alloy chemically reacts with dissolved oxygen according to Formula 3 to Formula 5.

(Formula 3)
Zr = Zr ⁴⁺ + 4e ⁻

(Formula 4)
O ₂ + 4e ⁻ = 2O ²⁻

(Formula 5)
Zr + O ₂ = ZrO ₂

  In an environment where there is no dissolved oxygen, zirconium chemically reacts with water according to Equation 3, Equation 6, and Equation 7.

(Formula 6)
^{_{H 2 O + 2e - = H}} 2 + O 2-

(Formula 7)
Zr + 2H ₂ O = ZrO ₂ + 2H ₂

  Depending on the temperature, the corrosion potential is as low as -1000 to -400 mV (SHE) due to the reaction of Equation 3. That is, due to the presence of zirconium, chromium ions present as chromate ions corroded and eluted from the heat transfer surface of the feed water heater can be reduced according to Equation 1 and further precipitated according to Equation 2. Zirconium is preferable from the viewpoint of radiation exposure because it has little effect on activation even if it enters the reactor. In addition to zirconium, the same effect can be obtained with titanium or hafnium. By removing chromium ions from the feed water and reducing the inflow of chromium ions into the reactor pressure vessel, the accumulation of chromium oxide on the surface of the reactor internal structure can be suppressed, and the chromium ion concentration at the time of stopping hydrogen injection can be reduced. An abrupt increase and, consequently, an increase in electrical conductivity can be suppressed.

  In the third method, during the period from the start of the plant shutdown operation until the reactor water temperature reaches 100 ° C., a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration is added to the reactor. Injected into water.

  By maintaining the reducing environment in the reactor, the elution of chromium ions from chromium oxide adhering to the surface of the reactor internal structure can be suppressed, and the chromium ion concentration rapidly increases when hydrogen injection is stopped. As a result, an increase in electrical conductivity can be suppressed.

  In the fourth method, a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration during the period from the start of the plant shutdown operation until the reactor water temperature reaches 100 ° C. Then, an oxidant is injected into the reactor water during a period when the reactor water temperature is 100 ° C. or lower.

  In the period from the start of the plant shutdown operation until the reactor water temperature reaches 100 ° C., by maintaining the reducing environment in the reactor, the chromium ions from the chromium oxide adhering to the surface of the reactor internal structure Suppresses elution. Furthermore, during the period when the reactor water temperature is 100 ° C or lower, oxidants such as oxygen, hydrogen peroxide, and ozone are added to the reactor water to make the inside of the reactor an oxidizing environment. This promotes the elution of chromium ions from the chromium oxide adhering to the surface of the reactor internal structure. It should be noted that the effective oxygen concentration is desirably 32 μmol / L or less when the oxidant is injected into the nuclear reactor. In an environment where the effective oxygen concentration is 32 μmol / L or less and the temperature of the reactor water is 100 ° C. or less, since the SCC sensitivity of stainless steel is small, the effect on the soundness of the plant is sufficiently small Because it is. When oxygen, hydrogen peroxide, and ozone are used as the oxidizing agent, the effective oxygen concentration can be expressed as shown in Equation 8.

(Formula 8)
[O ₂ ] eff = [O ₂ ] + (1/2) [H ₂ O ₂ ] + (2/3) [O ₃ ]
Here, [O ₂ ] eff is the effective oxygen concentration (mol / L), [O ₂ ] is the oxygen concentration (mol / L),
[H ₂ O ₂ ] is the hydrogen peroxide concentration (mol / L), and [O ₃ ] is the ozone concentration (mol / L).

  Thus, the amount of chromium oxide accumulated on the surface of the reactor internal structure can be reduced by injecting the oxidizing agent into the reactor water during the period when the reactor water temperature is 100 ° C. or less after the start of the plant shutdown operation. Therefore, since it is possible to suppress the elution of chromium ions during the next plant operation, it is possible to suppress a rapid increase in the chromium ion concentration and an increase in electrical conductivity when the hydrogen injection is stopped.

  In each of the above methods, hydrogen is used as a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration. However, in addition to hydrogen, ammonia, hydrazine, alcohol, etc. are used. Similarly to the case of hydrogen injection, chromium oxide accumulates on the surface of the reactor internal structure, and when the plant is shut down, the chromium ion concentration increases rapidly due to the elution of chromium ions, and the electrical conductivity increases. An increase is expected to occur. However, by using each of the above methods, even when ammonia, hydrazine, alcohol, or the like is used, a rapid increase in chromium ion concentration and an increase in electrical conductivity can be suppressed as in the case of hydrogen injection.

  A first embodiment of the stress corrosion cracking mitigation method for nuclear plant structural materials according to the present invention will be described below with reference to FIGS. 1 and 4 to 6. In this embodiment, a spinel structure compound layer composed of divalent metal ions and trivalent iron ions is formed on the heat transfer surface of the feed water heater based on the first method described above.

  FIG. 4 shows a boiling water nuclear plant to which the stress corrosion crack mitigation method of the first embodiment is applied. The boiling water nuclear power plant connects a condensate cooler 13, a condensate filtration demineralizer 3, a feed water pump 4, a feed water heater 5, and a nuclear reactor pressure vessel 1 loaded with nuclear fuel through a feed water piping 6. The reactor pressure vessel 1 and the turbine 2 are connected by the main steam pipe 14, and the turbine 2 and the condensate cooler 13 are connected to form a closed loop. Water is used as the reactor coolant. Water used as a coolant is called cooling water. Water is converted into steam in the reactor pressure vessel 1, the turbine 2 is rotated using the steam, and a generator (not shown) is rotated by the turbine 2 to generate power. The steam used to rotate the turbine 2 is returned to water by the condensate cooler 13, impurities are removed by the condensate filtration demineralizer 3, and then fed to the reactor pressure vessel 1 through the feed water heater 5 by the feed water pump 4. Returned. Separately, the reactor coolant recirculation system piping 16 connects the lower part of the reactor pressure vessel 1, the reactor coolant recirculation pump 7, and the jet pump 15 inlet. The heat output can be increased by increasing the flow rate of the coolant flowing through the reactor core by the reactor coolant recirculation pump 7. In the present embodiment, description will be made using a nuclear reactor having a reactor coolant recirculation system pipe 16. In this reactor, the reactor cooling water recirculation system piping 16 upstream side, the reactor cooling water purification system pump 9, the reactor cooling water purification system heat exchanger 11, the reactor cooling water filtration demineralizer 12, and the feed water system piping. 6 is connected by a reactor cooling water purification system pipe 10 and the reactor cooling water purification system pump 9 passes the cooling water to the reactor cooling water filtration demineralizer 12, thereby purifying impurities in the reactor water. To do. Also, a bottom drain pipe 8 that connects the bottom of the reactor pressure vessel 1 and the reactor cooling water purification system pipe 10 is installed. Further, an emergency core cooling system for injecting cooling water into the reactor core in order to cool the core in an emergency and a control rod for controlling the nuclear reaction of nuclear fuel are driven above the core of the reactor pressure vessel 1. A control rod drive hydraulic system for injecting cooling water into the tank is installed (not shown). Furthermore, the water quality in each system piping is monitored by the water quality monitors 21 to 25, and the dose rate of the main steam piping 14 is monitored by the main steam piping dose rate measuring device 26.

  Hydrogen is injected into the reactor water for the purpose of mitigating SCC of the in-reactor structure of the boiling water nuclear power plant. The injection of hydrogen into the reactor water is performed continuously or intermittently during the operation of the nuclear power plant. The period in which hydrogen can be injected is limited to the plant operation period in which the cooling water that has exited the reactor cooling water filtration demineralizer 12 is sent to the reactor pressure vessel 1. Thus, in order to inject hydrogen into the reactor water, a hydrogen gas generator 31 and a hydrogen gas injection amount adjustment valve 32 are installed in the water supply system pipe 6 in the water supply system pipe 6. Since the water supply system pipe 6 has a relatively low pressure, hydrogen can be easily injected into the cooling water. In this embodiment, hydrogen is injected for the purpose of mitigating SCC of the reactor internal structure. However, if the substance can be chemically reacted with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration, ammonia, Hydrazine, alcohol, etc. may be used.

  Next, a processing facility (hereinafter simply referred to as “processing facility”) for forming a compound layer of a spinel structure composed of divalent metal ions and trivalent iron ions on the surface of the heat transfer surface of the feed water heater. A procedure for forming a compound layer having a spinel structure on the heat transfer surface of the feed water heater during the shutdown period of the boiling water nuclear power plant will be described.

FIG. 6 shows a schematic structure of the processing facility. A valve 102, a surge tank 106, a temperature regulator 105, a pump 104 and a valve 117 are connected to the feed water heater 5 by a pipe 103 to form a closed loop. This closed loop allows chemicals to circulate between these devices and the feed water heater 5. The pipe 103 is provided with a water supply / drainage valve 107.
A pH adjusting chemical tank 110 is connected to the pipe 103 by a pipe 108, and a pump 109 and a valve 118 are connected to the pipe 108. A redox potential adjusting chemical tank 113 is connected to the pipe 103 by a pipe 111, and a pump 112 and a valve 119 are connected to the pipe 111. A metal ion chemical tank 116 is connected to the pipe 103 by a pipe 114, and a pump 115 and a valve 120 are connected to the pipe 114.

  By bringing the solution containing the metal ions constituting the spinel structure compound into contact with the feed water heater heat transfer surface, a spinel structure compound layer can be formed on the surface of the feed water heater heat transfer surface. . If necessary, the pH and oxidation-reduction potential are adjusted. Specifically, a spinel-type compound layer is formed on the surface of the feed water heater heat transfer surface by the following procedure.

  (S1) First, hydrogen injection into the reactor water is stopped during the plant shutdown period.

  (S2) Thereafter, the valves 102, 117, and 107 are opened, the valves 118, 119, and 120 are closed, and pure water is supplied from the valve 107 to the feed water heater 5 and the processing facility system.

  (S3) Next, the metal ion chemical solution tank 116 is supplied with a metal ion solution constituting a spinel-type compound composed of divalent metal ions and trivalent iron ions formed on the surface of the feed water heater heat transfer surface. To do. Iron, nickel, zinc, magnesium, and manganese can be used as a metal element used as a divalent metal ion. As the metal ion solution, a solution of a reagent selected from acetic acid, nitric acid, sulfate, and organic acid salt can be used. The organic acid can be converted to gas or water by decomposing with a catalyst or the like, and as a result, the amount of waste can be reduced. Therefore, an organic acid salt is more preferable as the metal ion solution.

  Further, a solution of a reagent that oxidizes and reduces iron ions is supplied to the oxidation-reduction potential adjusting chemical tank 113. Sodium nitrite, hydrogen peroxide, hydrazine, or the like can be used as a reagent solution for oxidizing and reducing iron ions. Further, a gas such as oxygen or ozone may be used. When using gas, the system water may be bubbled with a surge tank.

  Further, a reagent solution for adjusting the pH of the system water is supplied to the pH adjusting chemical tank 110. As the reagent solution for adjusting the pH, ammonia, sodium acetate, sodium hydroxide, hydrazine, or the like can be used.

  For these solutions, dissolved oxygen in the solution can be removed by bubbling with nitrogen or argon gas. In addition, you may implement supply operation | work of these solutions before S1.

  (S4) The system water is circulated by the pump 104 and the system water is heated to a predetermined temperature by the temperature regulator 105. The temperature of the system water is preferably high so that it does not boil, more preferably about 90 ° C., in order to promote the reaction.

  (S5) The valve 120 is opened, the pump 115 is started, and the metal ion solution constituting the compound having the spinel structure is injected into the processing facility system. This is because a solution containing metal ions constituting the spinel structure compound is brought into contact with the feed water heater heat transfer surface to form a spinel structure compound layer on the feed water heater heat transfer surface. When the system water reaches a predetermined concentration (system concentration is 1 to 100 mmol / L), the pump 115 is stopped, the valve 120 is closed, and the injection is stopped.

  (S6) The valve 118 is opened, the pump 109 is activated, and a reagent solution for adjusting the pH is injected into the processing facility system. When the system water reaches a predetermined pH (6 to 11 at room temperature pH), the pump 109 is stopped, the valve 118 is closed, and the injection is stopped.

  (S7) The valve 119 is opened, the pump 112 is started, and the Fe ions are oxidized and reduced so that the redox potential of the system water becomes a predetermined potential (−0.1 to −0.8 V (SHE)). The reagent solution to be injected is intermittently injected into the system water. In addition, when using gas, such as oxygen and ozone, what is necessary is just to carry out gas bubbling of system water with a surge tank.

  (S8) Then, it holds until the layer of the compound of the spinel type structure formed in the surface of a feed water heater heat-transfer surface becomes predetermined thickness (1 micrometer-170 micrometers).

Here, the feed water heater 5 uses a part of the steam from the reactor pressure vessel 1, and the cooling water exiting the condensate filtration demineralizer 3 is about 180-220 ° C. at the final stage outlet of the feed water heater. It is an apparatus which supplies to the reactor pressure vessel 1 after heating up to. Therefore, although the heat transfer efficiency of the feed water heater heat transfer surface is important, it is considered that the heat transfer efficiency is affected by the thickness of the oxide film formed on the surface. FIG. 5 shows the temperature dependence of the corrosion rate of SUS304 steel used for the feed water heater heat transfer surface. The horizontal axis in FIG. 5 indicates the temperature, and the vertical axis indicates the corrosion rate. The legend “Kobayashi et al.” Is the data described in “Corrosion rate measurement of stainless steel in high temperature water by AC impedance method” Corrosion Protection Lecture Proceedings (May 1984), “Honda et al. (1) "Is data described in" Boshoku Gijyutu, 37, p.287 (1988) "," Honda et al. (2) "is data described in" Anti-corrosion technology, 36, p.646 (1987) " is there. From Figure 5, the corrosion amount of about 80 mg / cm ² in the case of driving a 30-year nuclear power plant, about 180 mg / cm ² in the reactor water temperature of 200 ° C., at 180 ° C. to about 120mg / cm ^2, 160 ℃ below I understand that. Assuming that the density of the oxide film is 5.4 g / cm ³ which is the density of Fe ₃ O ₄ , the oxide film thickness is about 340 μm at 200 ° C., about 220 μm at 180 ° C., and about 150 μm at 160 ° C. or less. From the experimental results shown in FIG. 1, it can be considered that the corrosion rate of SUS304 steel provided with Fe ₃ O ₄ is about ½ that of untreated steel. It is thought that the heat transfer efficiency from the heat transfer surface of the heater can be maintained. That is, the thickness of the spinel structure compound layer formed on the surface of the heat transfer surface of the feed water heater is about 170 μm at a site of 200 ° C., about 110 μm at a site of 180 ° C., and about 75 μm at a site of 160 ° C. or less. The upper limit may be set. On the other hand, the spinel compound oxide formed on the surface of the heat transfer surface of the feed water heater was found from the experimental results of FIG. 1 that the corrosion rate could be suppressed by reducing the thickness of the spinel structure compound layer to 1 μm. It is considered that the thickness should be 1 μm or more.

  When the spinel-type compound is attached to the surface of the water heater heating surface, a specimen having the same material and surface condition as the surface of the water heater heating surface is immersed in the surge tank 108. During the adhesion treatment, the thickness of the spinel oxide formed on the surface of the heat transfer surface of the feed water heater can be determined by taking out this test piece in a timely manner and measuring the amount of weight change. That is, if the weight of the test piece at the time of removal, the weight of the initial test piece, and the surface area of the test piece are known, the thickness of the spinel oxide can be obtained from Equation 9.

(Formula 9)
(Thickness / μm) = {(Weight at extraction / g) − (Initial weight / g)} / (Surface area / cm ² )
× 1.85 × 10 ^-9

  (S9) When the spinel-type compound layer formed on the surface of the feed water heater heat transfer surface exceeds a predetermined thickness, the temperature controller 105 cools the system water to room temperature.

  (S10) Thereafter, the valve 107 is opened to drain the system water from the treatment facility. Pure water is injected into the system as necessary, and the feed water heater 5 is washed with water. Thus, all the steps of the method for mitigating stress corrosion cracking of the nuclear plant structural material according to this embodiment are completed.

  According to the present embodiment, by forming a spinel structure compound layer composed of divalent metal ions and trivalent iron ions on the feed water heater heat transfer surface, the feed water heater heat transfer surface with respect to the base material Corrosion can be suppressed, and elution of chromium ions accompanying corrosion can be suppressed. In other words, by suppressing the elution of chromium ions from the heat transfer surface of the feed water heater and reducing the inflow of chromium ions into the reactor pressure vessel, the accumulation of chromium oxide on the surface of the reactor internal structure is suppressed. can do. Therefore, even when hydrogen injection into the reactor water is stopped, an increase in the chromium ion concentration in the reactor water can be suppressed. As a result, an increase in electrical conductivity is suppressed, and as a result, stress corrosion. It becomes possible to suppress an increase in sensitivity to cracking.

  As a conventional technique, there is a method in which a dense oxide film is formed on the surface of the base material by oxidizing the surface of the base material in a high-temperature gas at 800 ° C. or higher during plant construction. However, it is very difficult to perform such treatment in a plant that has already started operation. On the other hand, since the method of forming a spinel structure compound on the heat transfer surface of the feed water heater during the plant shutdown period can be carried out at 100 ° C. or lower as in this embodiment, Also, it can be processed easily.

  The feed water heater 5 is composed of a plurality of devices, and is configured to increase the temperature step by step in each device from the upstream side. Since the amount of chromium ions generated increases in proportion to the amount of corrosion on the heat transfer surface of the feed water heater, if it is carried out on the high temperature side where the amount of corrosion on the heat transfer surface of the feed water heater is large, particularly 160 ° C or higher, The amount of chromium ions generated can be effectively suppressed.

  Next, a second embodiment of the stress corrosion cracking mitigation method for nuclear plant structural materials according to the present invention will be described with reference to FIG. In this embodiment, zirconium or a zirconium alloy is installed on the surface of the piping connecting the feed water heater and the pressure vessel that contacts the cooling water.

As shown in FIG. 7, a zirconium plate 201 assembled in a lattice shape so as to be parallel to the flow of cooling water is installed in a pipe 202. By installing the zirconium plate 201 assembled in a lattice shape in the pipe 202, the chromium ions corroded and eluted from the feed water heater are reduced and deposited on the zirconium plate 201.

  According to the present embodiment, chromium ions are removed from the feed water by precipitating chromium ions on the surface of the zirconium plate, and the inflow of chromium ions to the reactor pressure vessel is reduced. Oxide accumulation can be suppressed. Therefore, even when hydrogen injection into the reactor water is stopped, an increase in the chromium ion concentration in the reactor water can be suppressed. As a result, an increase in electrical conductivity is suppressed, and as a result, stress corrosion. It becomes possible to suppress an increase in sensitivity to cracking.

  In the above embodiment, the zirconium plate 201 is used in the pipe connecting the feed water heater and the pressure vessel. However, the same effect can be obtained by using a zirconium alloy plate instead of the zirconium plate 201. Instead of the zirconium plate 201, a plate obtained by cladding a thin film of zirconium can also be used. Furthermore, a zirconium alloy thin film may be directly clad on the water contact surface of the water supply system pipe. In addition, when using a plate clad with a zirconium thin film, the influence of the flow of cooling water is kept small compared to the case where a zirconium alloy thin film is directly clad on the water contact surface of the water supply system pipe, The liquid contact area between the cooling water and zirconium can be increased. The same effect can be obtained by cladding a zirconium alloy thin film instead of cladding a zirconium thin film.

  Next, a third embodiment of the stress corrosion cracking mitigation method for nuclear plant structural materials according to the present invention will be described with reference to FIG. This example is based on the third method described above, and during the period when the reactor water temperature is 100 ° C. or higher from the start of the plant shutdown operation, it chemically reacts with oxygen or hydrogen peroxide contained in the reactor water. A substance that reduces the concentration is injected into the reactor water.

  FIG. 8 shows a boiling water nuclear power plant to which the stress corrosion crack mitigation method of the third embodiment is applied. Since the nuclear power plant to which the third embodiment is applied has the same structure as the nuclear power plant shown in FIG. 4 described in the first embodiment, detailed description thereof will be omitted. However, it differs from the boiling water nuclear power plant described in the first embodiment in that a hydrogen gas generator 301 and a hydrogen gas injection amount adjustment valve 302 are connected to the reactor water purification system pipe 10.

  Below, during the period from the start of the plant shutdown operation to the reactor water temperature of 100 ° C., a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration is injected into the reactor water. The procedure is shown below. In this embodiment, hydrogen is used as a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration.

  First, the supply of cooling water from the water supply system pipe 6 is stopped. Thereafter, the hydrogen gas injection amount adjustment valve 302 is opened, and hydrogen is injected into the reactor water. Hydrogen can be injected into the reactor water continuously or intermittently. In addition, when hydrogen is injected into the reactor water, the dissolved oxygen concentration at the bottom of the reactor is measured by the water quality monitor 22. Based on the measurement result, the hydrogen injection amount is adjusted so that the dissolved oxygen concentration in the reactor water is approximately equal to or lower than the concentration during the period in which the cooling water is supplied from the water supply system. Thereafter, when the reactor water temperature becomes 100 ° C. or lower, the hydrogen gas injection amount adjusting valve 302 is closed to stop the hydrogen injection. Thus, all the steps of the method for mitigating stress corrosion cracking of the nuclear plant structural material according to this embodiment are completed.

  According to the present embodiment, elution of chromium ions from the chromium oxide attached to the surface of the in-reactor structure can be suppressed by maintaining the reducing environment in the reactor. Therefore, even when hydrogen injection into the reactor water is stopped, an increase in the chromium ion concentration in the reactor water can be suppressed. As a result, an increase in electrical conductivity is suppressed, and as a result, stress corrosion. It becomes possible to suppress an increase in sensitivity to cracking.

  In the present embodiment, hydrogen is taken as an example of a substance showing the effect of reducing oxygen or hydrogen peroxide contained in the reactor water, but a reducing agent such as ammonia or hydrazine or alcohol may be used.

  Further, the hydrogen gas generator 301 and the hydrogen gas injection amount adjustment valve 302 may be connected to the reactor cooling water recirculation system pipe 16 instead of the reactor water purification system pipe 10.

  Furthermore, in this embodiment, in order to control the dissolved oxygen concentration in the reactor water, the dissolved oxygen concentration at the bottom of the reactor was measured by the water quality monitor 22, but the chromium ion concentration was measured by the water quality monitor 22, and this result was obtained. It is also possible to control the dissolved oxygen concentration in the reactor water.

  Next, a fourth embodiment of the stress corrosion cracking mitigation method for nuclear plant structural materials according to the present invention will be described with reference to FIG. The present embodiment is based on the above-described fourth method and chemically reacts with oxygen or hydrogen peroxide contained in the reactor water during the period when the reactor water temperature is 100 ° C. or higher after the start of the plant shutdown operation. Then, a substance for reducing the concentration is injected into the reactor water, and then an oxidizing agent is injected into the reactor water during a period when the reactor water temperature is 100 degrees or less.

  FIG. 9 shows a boiling water nuclear plant to which the stress corrosion crack mitigation method of the fourth embodiment is applied. Since the nuclear power plant to which the fourth embodiment is applied has the same structure as the nuclear power plant shown in FIG. 8 described in the third embodiment, detailed description thereof will be omitted. However, it differs from the boiling water nuclear power plant described in the first embodiment in that a hydrogen peroxide tank 401 and a hydrogen peroxide injection pump 402 are connected to the reactor water purification system pipe 10. In this embodiment, hydrogen peroxide is used as the oxidizing agent, but oxygen or ozone may be used instead of hydrogen peroxide.

  In the following, substances that cause a chemical reaction with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration after the start of the plant shutdown operation and during which the reactor water temperature is 100 ° C. or higher are indicated in the reactor water. Then, a procedure for injecting an oxidant into the reactor water during a period when the reactor water temperature is 100 degrees or less will be described. In this embodiment, hydrogen is used as a substance that chemically reacts with oxygen or hydrogen peroxide contained in the reactor water to reduce its concentration.

  First, as in the third embodiment, during the period from the start of the plant shutdown operation to the reactor water temperature of 100 ° C., the chemical reaction with oxygen or hydrogen peroxide contained in the reactor water causes a reduction in the concentration. The material hydrogen is injected into the reactor water. Specifically, it is as follows. The supply of cooling water from the water supply system pipe 6 is stopped. Thereafter, the hydrogen gas injection amount adjustment valve 302 is opened, and hydrogen is injected into the reactor water. Hydrogen can be injected into the reactor water continuously or intermittently. In addition, when hydrogen is injected into the reactor water, the dissolved oxygen concentration at the bottom of the reactor is measured by the water quality monitor 22. Based on the measurement result, the hydrogen injection amount is adjusted so that the dissolved oxygen concentration in the reactor water is approximately equal to or lower than the concentration during the period in which the cooling water is supplied from the water supply system. Hydrogen is injected into the reactor water during a period in which the reactor water temperature is up to 100 ° C. Thereafter, the hydrogen gas injection amount adjusting valve 302 is closed to stop hydrogen injection.

  Subsequently, when the reactor water temperature becomes 100 ° C. or lower, the hydrogen peroxide solution is injected from the hydrogen peroxide solution tank 401 into the reactor water by the hydrogen peroxide injection pump 402. The hydrogen peroxide solution can be injected into the reactor water continuously or intermittently. In addition, when the hydrogen peroxide solution is injected into the reactor water, the dissolved oxygen concentration at the bottom of the reactor is measured by the water quality monitor 22. Based on the measurement result, the injection amount of the hydrogen peroxide solution is adjusted so that the effective oxygen concentration is 32 μmol / L or less. When the amount of chromium oxide accumulated in the structural material is sufficiently reduced, the hydrogen peroxide injection pump 402 is stopped and injection of hydrogen peroxide into the reactor water is stopped. The chromium ion concentration increases when hydrogen peroxide injection is started, and decreases when the chromium oxide accumulated in the structural material is sufficiently reduced. Therefore, by monitoring the chromium ion concentration, it can be determined whether or not the amount of chromium oxide accumulated in the structural material has been sufficiently reduced. For example, the time point when the chromium ion concentration becomes 1/10 or less of the maximum value can be considered as the chromium ion concentration when the amount of chromium oxide accumulated in the structural material is sufficiently reduced. Thus, all the steps of the method for mitigating stress corrosion cracking of the nuclear plant structural material according to this embodiment are completed.

  According to this embodiment, as in the third embodiment, during the period from the start of the plant shutdown operation to the reactor water temperature of 100 ° C., it chemically reacts with oxygen or hydrogen peroxide contained in the reactor water. By injecting substances that reduce the concentration into the reactor water, it is possible to maintain the reducing environment in the reactor and to suppress the elution of chromium ions from the chromium oxide attached to the surface of the structure inside the reactor. . Furthermore, the amount of chromium oxide accumulated on the surface of the reactor internal structure can be reduced by injecting an oxidant into the reactor water after the plant shutdown operation has started and the reactor water temperature is 100 ° C or lower. The elution of chromium ions during the next plant operation can be suppressed. Therefore, even when hydrogen injection into the reactor water is stopped, an increase in the chromium ion concentration in the reactor water can be suppressed. As a result, an increase in electrical conductivity is suppressed, and as a result, stress corrosion. It becomes possible to suppress an increase in sensitivity to cracking.

  In this embodiment, the hydrogen peroxide solution tank 401 and the hydrogen peroxide injection pump 402 are connected to the reactor water purification system piping 10, but instead of the reactor water purification system piping 10, the reactor cooling water recirculation system piping 16 may be connected.

  In each of the above embodiments, the plant operation period is a period during which cooling water is injected from the water supply system into the reactor pressure vessel 1, and the plant stop period is a period during which all control rods are inserted and the reactor The period during which the lid of the pressure vessel is opened, the plant shutdown operation period is the period from the plant operation period until the cooling water injection from the feed water system to the reactor pressure vessel 1 is stopped and the control rods are fully inserted. A period.

Fe ₃ O ₄ shows the results of corrosion tests of SUS304 steel coating was formed. The figure which shows the relationship between CGR and ECP of SUS304 steel. The figure which shows the dependence of ECP of SUS304 steel with respect to oxygen or hydrogen peroxide concentration. The figure which shows the boiling water nuclear power plant which applies the stress corrosion crack mitigation method of a 1st Example. The figure which shows the temperature dependence of the corrosion rate of SUS304 steel. The figure which shows the processing equipment for forming the layer of the compound of a pinel type structure on the surface of a feed water heater heat-transfer surface. The figure which shows the boiling water nuclear power plant which applies the stress corrosion crack mitigation method of a 2nd Example. The figure which shows the boiling water nuclear power plant to which the stress corrosion crack mitigation method of the 3rd Example is applied. The figure which shows the boiling water nuclear power plant to which the stress corrosion crack mitigation method of the 4th Example is applied.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Reactor pressure vessel, 5 ... Feed water heater, 31 ... Hydrogen gas generator, 110 ... pH adjustment chemical liquid tank, 113 ... Redox potential adjustment chemical liquid tank, 116 ... Metal ion chemical liquid tank, 201 ... Zirconium plate, 301 ... Hydrogen gas generator 401, hydrogen peroxide water tank.

Claims (10)

During operation of a boiling water nuclear plant, a substance that chemically reacts with oxygen or hydrogen peroxide to reduce the concentration of oxygen or hydrogen peroxide in the reactor water is injected into the reactor water of the boiling water nuclear plant. And
During a shutdown period of the boiling water nuclear plant, a compound layer having a spinel structure composed of divalent metal ions and trivalent iron ions is formed on the heat transfer surface of the feed water heater of the boiling water nuclear plant. To reduce stress corrosion cracking of nuclear plant structural materials.   The formation of the spinel structure compound layer on the feed water superheater heat transfer surface according to claim 1, wherein the spin water structure heat transfer surface includes a solution containing metal ions constituting the spinel structure compound. Stress corrosion cracking mitigation method for nuclear plant structural materials by contact.   In Claim 1 or 2, formation of the layer of the compound of the spinel structure on the heat transfer surface of the feed water superheater constitutes the compound of the spinel structure in system water in a closed loop including the feed water heater. A method for mitigating stress corrosion cracking of a nuclear plant structural material, which is performed by injecting a solution containing metal ions, a reagent for adjusting the pH of the system water, and a reagent for adjusting a redox potential of the system water.   4. The method according to claim 1, wherein the formation of the compound layer having the spinel structure on the heat transfer surface of the feed water superheater is performed during a shutdown period of the boiling water nuclear power plant. Is a method for mitigating stress corrosion cracking of nuclear plant structural materials performed during a period of 100 degrees or less.   5. The stress corrosion cracking mitigation of a nuclear plant structural material according to claim 1, wherein the divalent metal ion is one or more metal ions selected from iron, nickel, zinc, magnesium and manganese. Method. During operation of a boiling water nuclear plant, a substance that chemically reacts with oxygen or hydrogen peroxide to reduce the concentration of oxygen or hydrogen peroxide in the reactor water is injected into the reactor water of the boiling water nuclear plant. And
Stress corrosion cracking mitigation method for nuclear plant structural material that supplies cooling water into the pipe in a state where zirconium or zirconium alloy is installed in the pipe connecting the water heater and pressure vessel of the boiling water nuclear plant .   7. The method for mitigating stress corrosion cracking of a nuclear plant structural material according to claim 6, wherein the zirconium or zirconium alloy is a zirconium plate or a zirconium alloy plate assembled in a lattice shape. During operation of a boiling water nuclear plant, a substance that chemically reacts with oxygen or hydrogen peroxide to reduce the concentration of oxygen or hydrogen peroxide in the reactor water is injected into the reactor water of the boiling water nuclear plant. And
After the start of the shutdown operation of the boiling water nuclear power plant and during a period when the temperature of the reactor water is 100 ° C. or higher, the concentration of oxygen or hydrogen peroxide in the reactor water is set in the reactor water. A method for mitigating stress corrosion cracking in nuclear plant structural materials by injecting substances to be reduced. During operation of a boiling water nuclear plant, a substance that chemically reacts with oxygen or hydrogen peroxide to reduce the concentration of oxygen or hydrogen peroxide in the reactor water is injected into the reactor water of the boiling water nuclear plant. And
The concentration of oxygen or hydrogen peroxide in the reactor water is reduced in the reactor water after the start of shutdown operation of the boiling water nuclear plant and the temperature of the reactor water is 100 ° C. or higher. Injecting the substance
Stress corrosion cracking mitigation of nuclear plant structural material into which oxidant is injected into the reactor after the start of the shutdown operation of the boiling water nuclear plant and the temperature of the reactor water is 100 ° C. or less Method. The stress corrosion cracking mitigation method for a nuclear plant structural material according to any one of claims 1 to 9, wherein the substance is at least one of hydrogen, ammonia, hydrazine, and alcohol.

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