JP4717388B2 - Hydrogen injection method for boiling water nuclear power plant - Google Patents

Description

  The present invention relates to a hydrogen injection method used for mitigating a corrosive environment in a nuclear reactor of a boiling water nuclear power plant, and more particularly to a method of performing hydrogen injection when starting a boiling water nuclear power plant.

  When stress is applied to a metal material in a corrosive environment, the metal material may be broken at a lower stress than when the metal material is not in a corrosive environment. This phenomenon is called stress corrosion cracking (hereinafter referred to as “SCC”). In-furnace structures and piping structural materials may cause SCC in a high-temperature water environment. SCC is a material factor (ex. Material sensitization: Phenomenon in which corrosion resistance deteriorates due to a chromium-deficient layer at the grain boundary due to the thermal effect of welding), stress factor (ex. Presence of welding residual stress, etc.), environment This occurs when three factors (ex. Presence of dissolved oxygen, etc.) overlap. Therefore, the occurrence of SCC can be prevented by eliminating one or more of the above three factors.

  Hydrogen injection (Hydrogen Water Chemistry, hereinafter referred to as “HWC”) as one of the SCC countermeasures for the structural material in the reactor of boiling water nuclear power plant (hereinafter referred to as “BWR (Boiling Water Reactor) power plant”) There is. HWC is a technology that reduces the concentration of oxidants such as dissolved oxygen and hydrogen peroxide, which are SCC environmental factors, by suppressing the radiolysis of water by hydrogen injected into the furnace and promoting recombination. It is applied to BWR power plant.

In general, hydrogen injection is performed by injecting hydrogen gas from a water supply system of a BWR power plant and feeding the hydrogen into the reactor in a state where hydrogen is dissolved in the water supply (see, for example, Patent Document 1).
JP-A-6-167596

  As described above, in the BWR power plant, hydrogen injection is performed as a measure for suppressing the occurrence and progress of SCC. However, conventional hydrogen injection is performed during normal operation of the BWR power plant, and no hydrogen is injected when the BWR power plant is started. The reason is as follows.

  (1) While the period of normal operation of the BWR power plant extends over a long period of about 12 months, the period required to start up after stopping the BWR power plant by periodic inspections is only a few days. This is because even if hydrogen injection is performed, the SCC suppression effect added by the hydrogen injection is considered to be small. In addition, because plant parameters such as reactor power, temperature, and pressure fluctuate during startup, it is difficult to inject hydrogen under appropriate control.

  (2) The hydrogen injection during normal operation generally employs a method of injecting hydrogen from the water supply system of the BWR power plant. However, at the time of start-up, this water supply system is in a standby state, water is not circulated from the water supply system into the reactor, and hydrogen cannot be injected into the core.

  As shown in FIG. 8, when the BWR nuclear power plant is started, the water flow rate in the feed water system does not increase until a predetermined time has elapsed after the BWR power plant is started (A in the figure). B) in the figure, a period in which hydrogen injection cannot be performed occurs.

  Therefore, if hydrogen injection is to be performed at start-up, it is necessary to prepare a separate system in which water is circulated into the reactor at start-up. It is difficult to install a separate system in the BWR power plant in various aspects such as cost and securing the installation location.

  However, during normal operation of the BWR power plant, the temperature and pressure in the reactor are almost constant, so that an environment in which a dynamic load (strain rate) is generated in the structural material in the reactor does not occur. On the other hand, when a BWR power plant is started up, it is estimated that a dynamic load (strain rate) is generated that may cause stress corrosion cracking in the structural material in the reactor due to the temperature rise and pressure increase in the reactor, although it is a short period of time. The

  Therefore, the in-reactor structural material at the start-up of the BWR power plant is placed in an environment where SCC is more likely to occur than during normal operation. In addition, when hydrogen injection is not applied (Normal Water Chemistry, hereinafter referred to as “NWC”), the concentration of an oxidizing agent such as hydrogen peroxide that easily causes SCC in the reactor water is low. It may increase remarkably by radiolysis, and can be said to be an environment in which SCC is likely to occur among environmental factors among the three SCC factors.

  Therefore, it is desirable to take measures to prevent stress corrosion cracking of the structural material in the reactor even when the BWR power plant is started up.

  The present invention has been made to solve such a conventional problem, and an object of the present invention is to provide a hydrogen injection method for a BWR power plant that suppresses the occurrence of SCC.

The invention according to claim 1 of the present invention relates to a reactor internal material in contact with reactor water of a boiling water nuclear power plant that reaches the rated output by repeating the boiling and non-boiling states of the reactor water during startup . A method for hydrogen injection in a boiling water nuclear power plant that suppresses stress corrosion cracking, wherein hydrogen is injected into the reactor water when the boiling water nuclear power plant is started up, and hydrogen is injected into the reactor water when the reactor water is in a boiling state. Injection of boiling water nuclear power plant characterized in that the amount of hydrogen injected is higher than the amount of hydrogen injected into the reactor water when the reactor water is in a non-boiling state to reduce the oxidant concentration in the reactor water Is the method.

  According to the hydrogen injection method of the present invention, it is possible to create an in-reactor environment in which stress corrosion cracking of a nuclear reactor structural material is unlikely to occur by performing hydrogen injection into the reactor at the time of severe start-up of the stress corrosion cracking environment. .

  Hereinafter, a BWR power plant according to an embodiment of the present invention will be described in detail with reference to the drawings.

  FIG. 1 is a diagram showing each system of the BWR power plant of this embodiment. The arrows in the figure indicate the direction in which the reactor water flows.

  The present inventors have thought that not only during normal operation of the BWR power plant, but also at the start-up, hydrogen injection may improve the environment and have the effect of extending the period until SCC occurs. It was. Therefore, a BWR power plant was constructed so that hydrogen could be injected into the reactor water even at startup. Details are as follows.

  The BWR power plant includes a reactor pressure vessel 1 and a turbine 2. A reactor core 4 is provided inside the reactor pressure vessel 1, and a reactor core shroud, a shroud support, a reactor core structural material 7 such as a reactor core support is installed so as to surround the reactor core 4, and a control rod (not shown) A fuel assembly is placed in the core 4. And the nuclear reactor structural material 7 is comprised with metal materials, such as stainless steel and a nickel base alloy.

  The reactor water supplied to the core 4 is heated by the fission reaction of the fuel assembly and becomes steam. This steam is guided to the turbine 2 by the main steam pipe 31. The turbine 2 is driven by this steam and rotates a generator (not shown) connected to the turbine 2 to generate power.

  Further, from the middle of the main steam pipe 31, the turbine bypass pipe 42 branches, and by opening and closing the turbine bypass valve 103 provided in the middle of the turbine bypass pipe 42, the steam does not pass through the turbine 2, It is supplied to the condenser 3.

  The steam discharged from the turbine 2 is condensed in the condenser 3 and supplied as water through the pump 27 in the water supply system 55, the feed water heater (CD) 11, the pump 26, the pump 25, and the valve 101 in this order. Water is supplied from the pipe 32 to the reactor pressure vessel 1.

  The reactor water in the reactor pressure vessel 1 is guided into the purification system pipe 39 of the purification system (CUW system) 54 by driving the pump 24. One end of the purification system pipe 39 is connected to the recirculation system pipe 40, and the other end is connected to the water supply pipe 32. The purification system pipe 39 is provided with a regenerative heat exchanger (regenerated Hx) 12, a non-regenerative heat exchanger (non-regenerated Hx) 13, and a desalter (F / D) 14. ing. The reactor water in the purification system 54 flows through the regenerative heat exchanger 12, the non-regenerative heat exchanger 13, the desalter (F / D) 14, and the regenerative heat exchanger 12 in this order, and is purified and supplied to the water supply pipe 32. Return to the reactor pressure vessel 1.

  Further, purification system sampling pipes 62 and 63 are provided near the inlet and the outlet of the desalter 14 in the middle of the purification system pipe 39. Water quality can be inspected by introducing the reactor water to the water quality measuring instrument (not shown) through the purification system sampling pipes 62 and 63. A CUW blowdown pipe 35 branches from the purification system pipe 39 and the valve 102 is opened, so that the reactor water in the purification system pipe 39 is supplied to the condenser 3.

  The recirculation system (PLR system) 51 includes a recirculation pump 21 and a recirculation system pipe 40. Reactor water in the reactor pressure vessel 1 returns to the reactor pressure vessel 1 through the recirculation piping 40 by driving the recirculation pump 21.

  A recirculation sampling pipe 61 is provided in the middle of the recirculation pipe 40. Water quality can be inspected by introducing the reactor water to a water quality measuring instrument (not shown) through the recirculation system sampling pipe 61.

  The residual heat removal system 52 is a system that removes decay heat from the core 4, and includes the pump 22, the heat exchanger 16, and the residual heat removal system piping 41. The pump 22 is driven, and the reactor water in the reactor pressure vessel 1 is sent to the heat exchanger 16 through the residual heat removal system piping 41, and after cooling, the reactor water is passed through the residual heat removal system piping 41 and inside the reactor pressure vessel 1 Return to.

  The control rod drive water system (CRD system) 53 includes a pump 23 and a control rod drive water pipe 36, and drives a control rod (not shown) with the drive water in the water system to control the output of the nuclear reactor.

  The oxygen injection system 6 is provided at the outlet of the air extractor 15 of the offgas system pipe 64 connected to the condenser 3. In the oxygen injection system 6, a predetermined amount of oxygen is injected into the off-gas system pipe 64, and the hydrogen injected from the hydrogen injection system 5 is recombined.

  The hydrogen injection system 5 is provided in connection with a purification system pipe 39 of a purification system (CUW system) 54. The hydrogen injection system 5 includes a valve 501 and a hydrogen supply unit 502. The hydrogen supply unit 502 is composed of, for example, a hydrogen cylinder. By opening and closing the valve 501, the hydrogen gas in the hydrogen cylinder is injected into the reactor water in the purification system 54. Then, by controlling the opening / closing operation of the valve 501, the hydrogen injection amount can be controlled.

  In the hydrogen injection method of the present embodiment, the hydrogen injection system 5 is provided in the middle of the purification system pipe 39 in the purification system 54. However, the installation location is not limited to this, and the reactor water continues during startup. Any system with a general circulation may be used. In addition to the purification system (CUW system) 54, for example, there are a control rod drive water system (CRD system) 53 and a recirculation system (PLR system) 51 as systems having continuous circulation of reactor water at the time of startup.

  However, when considering the amount of system water, influence on equipment, controllability (limitation of injection flow rate), construction scale, etc., the outlet of the desalinator (FD) 14 of the purification system 54 (downstream of the desalter 14) The position on the purification system pipe 39) is suitable as a hydrogen injection point.

  FIG. 2 shows the result of comparative examination of the injection position. Note that the “*” mark in the column indicating whether or not there is an influence on the device indicates a position where hydrogen accumulation may occur.

  As can be seen from FIG. 2, the outlet of the desalinator (FD) 14 of the purification system (CUW system) 54 has sufficient continuous circulation of the reactor water (system flow rate) even when the reactor is started up. Because of its high hydrogen solubility, there is little impact on downstream equipment due to the generation of hydrogen pools, etc., the amount of hydrogen injection can be easily controlled, and the construction scale when remodeling the power plant can be reduced. It is suitable as an installation location of the hydrogen injection system 5. An outlet of the recirculation system (PLR system) 51 is also suitable as an installation location of the hydrogen injection system 5.

  Further, a bottom drain line 65 is provided at the lower part of the reactor pressure vessel 1 to allow water intake from the reactor pressure vessel 1. And water quality can be inspected by introducing the reactor water to a water quality measuring instrument (not shown).

  Next, the operation at the time of start-up of the BWR power plant having the above configuration will be described mainly focusing on the temperature rise change of the reactor water.

  FIG. 3 is a diagram showing a temperature change of the reactor water in the reactor pressure vessel 1 of the BWR power plant of the present embodiment.

The vertical axis represents the temperature of the reactor water (° C.) and the amount of hydrogen (Nm ³ / hr) injected into the reactor water. The horizontal axis indicates the elapsed time (hr) from the start of the start of the BWR power plant. The dotted line indicates the change in the reactor water temperature, and the solid line indicates the change in the hydrogen injection amount.

  It takes about 3.5 days at the time of start-up until the rated state of normal operation is reached after the start of start-up. During this time, the temperature and pressure of the reactor water in the reactor pressure vessel 1 are strictly managed by a control device (not shown).

  The start-up time of the BWR power plant refers to a period from the start of the start-up of the nuclear reactor until the normal operation state is reached, and this period can be divided into the following four stages. That is, the degree of vacuum of the main condenser before pulling out the control rod is achieved (first stage), the temperature rise / boosting period of the reactor water by pulling out the control rod (second stage), the turbine startup / generator combination (third stage), And the output rise to the rated output (fourth stage). The output of the BWR power plant is increased by pulling out the control rod until the second stage, and by pulling out the control rod and increasing the speed of the recirculation pump 21 (that is, increasing the reactor water flow rate) after the third stage. In the meantime, the equipment necessary for the operation of the nuclear reactor and the turbine generator is started at an appropriate timing according to the increase in the reactor power or the reactor pressure.

  Here, the second stage, which is the temperature rise / pressurization period of the reactor water, proceeds as follows. First, the reactor mode switch is activated, and then a control rod (not shown) is gradually pulled out from the core 4 in accordance with a predetermined operation sequence.

  When the control rod is withdrawn and the reactor reaches criticality, the reactor power gradually increases. Control the operation of the control rod while monitoring the temperature rise rate of the reactor water below the specified value so that excessive thermal stress is not generated in the reactor pressure vessel 1 when the temperature of the reactor water (inside) is increased. To increase the reactor power. In this way, the system up to the reactor pressure vessel 1 and the turbine 2 inlet is heated up to the rated temperature. This temperature increase / pressurization period is about one day.

  In this startup, the reactor water is continuously circulated to the reactor pressure vessel 1 in the purification system 54, the control rod drive water system 53, and the recirculation system 51. There is no water circulation.

  Here, as shown in FIG. 3, the actual temperature of the reactor water rises by repeating a rise and a substantially constant state, and reaches the rated temperature during normal operation.

  This is because the start-up operation of the BWR power plant is performed so as to be constant at a predetermined reactor water temperature in advance. Since the BWR power plant uses the water-steam system in the furnace in a saturated state, the temperature and pressure in the furnace have a one-to-one correspondence. In the BWR power plant, the temperature of the reactor water is not directly controlled, and when the pressure corresponding to the predetermined reactor water temperature is reached, the pressure is increased by opening the turbine bypass valve 103 and releasing the steam. A constant value is set. And the control corresponding to the pressure and the temperature corresponding to one to one is performed.

  Under the condition where the temperature of the reactor water is substantially constant, the boiling phenomenon of the reactor water occurs in the reactor pressure vessel 1. Therefore, a considerable part of the energy by the nuclear heating is consumed for the boiling of the reactor water, does not contribute to the temperature rise of the reactor water, and the temperature of the reactor water becomes substantially constant. When the boiling phenomenon converges, the energy from the nuclear heating is consumed again to increase the reactor water temperature, and the reactor water temperature rises. Thereafter, this constant state and the rising state are repeated.

FIG. 3 shows the change in the hydrogen injection amount (Nm ³ / hr) at the time of startup. As described above, hydrogen is injected into the reactor water from the hydrogen injection system 5 provided at the outlet of the demineralizer 14 of the purification system 54. The valve 501 is operated to inject hydrogen from the hydrogen supply unit 502.

  Here, the dissolved hydrogen concentration in the reactor water is preferably 50 ppb or more. This is because when the dissolved hydrogen concentration in the reactor water is 50 ppb or more, the oxidation environment of the reactor water is relaxed and the occurrence of stress corrosion cracking in the reactor internal structure is suppressed.

  Conversely, if the dissolved hydrogen concentration in the reactor water is less than 50 ppb, the oxidant concentration may not necessarily be sufficiently low (the oxidant concentration is 1 ppb or less).

  From the viewpoint of reducing the oxidant concentration, there is no particular limitation on the upper limit of the dissolved hydrogen concentration, but actually it is preferably 110 ppb or less. When the dissolved hydrogen concentration exceeds 110 ppb, the hydrogen solubility in water is sufficient and there is no safety problem, but excessive hydrogen injection occurs. That is, the effect of reducing the oxidant concentration is saturated and there is a problem that hydrogen is wasted. Therefore, it is desirable that the dissolved hydrogen concentration in the reactor water be 110 ppb or less.

  Note that the measurement position of dissolved hydrogen in the reactor water is the purification system sampling pipe 63. System water is taken out from the purification system sampling pipe 63 provided before entering the filter, and measurement is performed. In addition, the dissolved hydrogen amount is measured by using a dissolved hydrogen meter having a thermal conductivity detection type detector using a gas separation membrane, connecting this dissolved hydrogen meter to the purification system sampling pipe 63, and passing the system water through. To measure.

  Hydrogen is injected into the reactor water from the hydrogen injection system 5 so that the dissolved hydrogen concentration in the reactor water is 50 ppb or more (preferably 50 to 110 ppb). The hydrogen injection amount is an amount estimated to be within the above concentration range. However, it is preferable that the actual hydrogen injection amount is added with a moderate margin.

The hydrogen injection amount in the BWR power plant of this embodiment will be described below. As shown in FIG. 3, the amount of hydrogen injected into the reactor water is about 2.5 Nm ³ / hr when the reactor water temperature is in an elevated state. On the other hand, when the reactor water temperature is constant aspect, from about 7.5 nm ³ / hr (when the reactor water temperature 190 ° C.), (when the reactor water temperature 250 ° C.) to about 12Nm ³ / hr, about 17 Nm ³ / hr (when the reactor water temperature is 280 ° C.) and the case where the reactor water temperature is rising, the injection amount is increased, and the injection amount is increased as the reactor water temperature is increased.

  This is because when the reactor water temperature is in a constant state, the reactor water is in a boiling state in the reactor pressure vessel 1, and hydrogen dissolved in the reactor water escapes to the steam layer along with this boiling. This is because hydrogen escapes from the off-gas piping 64 to the outside of the plant via the turbine 2 in response to opening and closing of the bypass valve 103, so that the amount of escape of hydrogen is compensated and the concentration of dissolved hydrogen is maintained in a range of 50 ppb or more.

  Thus, by changing the amount of hydrogen injected into the reactor water according to the boiling of the reactor water, the dissolved hydrogen concentration in the reactor water can be kept in the range of 50 ppb or more. In addition, in order to finely adjust the hydrogen injection amount in anticipation of the amount of hydrogen consumed by boiling the reactor water, it is no longer necessary to inject more hydrogen into the reactor water, resulting in less hydrogen consumption. It is possible to save.

  The control of the hydrogen injection amount may be performed by setting the hydrogen injection amount in advance based on the reactor water temperature rise curve set in advance during the start-up operation of the BWR power plant. The temperature and boiling state of the water may be sequentially measured, and the hydrogen injection amount may be controlled based on the measurement result. The above hydrogen injection amount is an example, and the hydrogen injection amount for maintaining the dissolved hydrogen concentration in the reactor water at 50 ppb or more is increased or decreased depending on the scale and design of the BWR power plant.

  As described above, according to the hydrogen injection method of the boiling water nuclear power plant of the present embodiment, hydrogen is injected into the reactor water even when the boiling water nuclear power plant, which is an environment in which SCC is likely to occur, is started. As a result, the oxidant concentration in the reactor water decreases, and it is possible to create an in-reactor environment in which stress corrosion cracking of the reactor structural material is unlikely to occur. Thereby, generation | occurrence | production of SCC of a nuclear reactor structural material can be suppressed effectively.

  Then, by injecting hydrogen from the purification system 54 with circulation of reactor water at the time of starting the boiling water nuclear power plant, the reactor water in which hydrogen is dissolved in the nuclear reactor structural material can be distributed even at the time of startup.

  Moreover, the oxidant concentration in the reactor water can be effectively reduced by injecting hydrogen so that the dissolved hydrogen concentration in the reactor water is 50 ppb or more, preferably 50 to 110 ppb.

  Then, by making the amount of hydrogen injected into the reactor water at the time of boiling the reactor water larger than the amount of hydrogen injected into the reactor water when there is no reactor water, the dissolved hydrogen concentration in the reactor water is kept within a predetermined range. It can be maintained and the consumption of hydrogen can be suppressed.

  The embodiment described above can be implemented with various changes and improvements.

  Examples of the present invention will be described below.

  FIG. 4 is a diagram illustrating a result of an experiment for confirming the effect of suppressing the occurrence of SCC by performing hydrogen injection at startup. The vertical axis of the graph shows an index (hazard function) of the possibility of occurrence of SCC, and the horizontal axis shows the test time (hr).

  Since it is difficult to conduct an experiment of SCC generation with an actual machine of a BWR power plant, an environment simulating the environment of the actual machine was created and a simulation experiment was conducted.

  An outline of the experiment is briefly explained. The test piece used in the experiment was obtained by subjecting stainless steel (SUS304) to sensitization (a heat treatment that intentionally causes stress corrosion cracking).

  The repeated load conditions on the test piece are (i) increasing the load at a constant rate over 3 hours from 3 Sm to 4 Sm. (Ii) The load is then rapidly returned to 3 Sm. Thereafter, the loads of (ii) and (ii) are repeated. Note that Sm is a unit representing the stress intensity for design defined by the structure standard.

  The repeated temperature conditions for the test piece were (i) 50 ° C. for 5.3 hours, (ii) 16 hours thereafter, increasing the temperature to 288 ° C. at a constant rate, and (iii) 8 hours. The temperature is lowered at a constant rate of decrease to 50 ° C. Thereafter, the temperature conditions (ii) and (iii) are repeated.

  Then, a group of materials that is an environment in which hydrogen is injected so that the dissolved hydrogen concentration of the reactor water becomes 50 ppb or more while applying dynamic strain to the structural material in the reactor (hereinafter referred to as “startup HWC”). And a start-up simulation test of the BWR power plant divided into a group of materials designated as an environment without hydrogen injection (hereinafter referred to as “start-up NWC”).

  As can be seen from FIG. 4, it was found that the fracture time due to SCC was 30 times longer in the group of materials subjected to hydrogen implantation than in the group of materials not subjected to hydrogen implantation. It has been found that injecting sucrose has a very large effect in suppressing the occurrence of SCC.

Next, FIGS. 5, 6, and 7 show the results of measuring the water quality in the reactor pressure vessel 1 when the hydrogen is not injected into the reactor water at startup (NWC at startup) and when it is performed (HWC at startup). The environmental factor is one of the three factors). In all cases, the temperature rise (° C.) of the reactor water and the hydrogen injection amount (Nm ³ / hr) at the start-up operation are the same as those shown in FIG.

FIG. 5 is a diagram showing the chromate ion (CrO ₄ ²⁻ ) concentration contained in the reactor water in the reactor pressure vessel when the BWR power plant is started up.

The vertical axis shows the temperature of the reactor water (° C.), the amount of hydrogen injected into the reactor water (Nm ³ / hr), and the chromate ion concentration (ppb). The horizontal axis indicates the elapsed time (hr) from the start of the start of the BWR power plant.

  The measurement position of the chromate ion concentration is the purification system sampling pipe 63. In addition, the chromate ion concentration measurement method uses a column packed with an ion-exchange resin to separate each ion and measure the conductivity of the ion component to purify an ion chromatograph that purifies the ion sampling system. It connects to 63 and it measures by passing system water.

  In an environment where SCC occurs, chromate ions elute into the reactor water from stainless steel, which is the material that mainly constitutes the structural material 7 in the reactor. Therefore, the chromate ion concentration in the reactor water is capable of generating SCC. It is an index of sex.

  As shown in FIG. 5, the chromate ion concentration in the start-up HWC (the line marked by the black square in the figure) is significantly higher than that in the start-up NWC (the line marked by the white square in the figure). It turns out that it has fallen (almost 0 ppb). That is, it shows that the corrosion of the in-reactor structural material 7 is suppressed by the startup HWC.

  Therefore, it can be seen that performing hydrogen injection at the start-up of the BWR power plant has a great effect on reducing the occurrence of SCC.

FIG. 6 is a diagram showing the concentrations of dissolved oxygen (DO) and hydrogen peroxide (H ₂ O ₂ ), which are oxidants, contained in the reactor water in the reactor pressure vessel when the BWR power plant is started up. .

The vertical axis represents the temperature of the reactor water (° C.), the amount of hydrogen injected into the reactor water (Nm ³ / hr), and the oxidant concentration (ppb). The horizontal axis indicates the elapsed time (hr) from the start of the start of the BWR power plant.

  The measurement position of the oxidant concentration is the purification system sampling pipe 63. The measuring method of the oxidant concentration is to measure the dissolved oxygen concentration by connecting a detector to the purification system sampling pipe 63 using a dissolved oxygen meter having a diaphragm type electrochemical detector and passing the system water. In addition, the hydrogen peroxide concentration is measured by thiocyanic acid method using a spectrophotometer by transporting the system water collected in the sample container from the purification system sampling pipe 63 to the analysis chamber.

The reactor water is decomposed by radiation radiated from the core 4, and dissolved oxygen (DO) and hydrogen peroxide (H ₂ O ₂ ) are generated in the reactor water. Since the crystal grain boundary of the reactor structural material 7 is corroded and eluted by this oxidant, the oxidant concentration in the reactor water is an index of the possibility of occurrence of SCC.

  As shown in FIG. 6, in the startup NWC, the dissolved oxygen concentration (solid line in the figure) and the hydrogen peroxide concentration (broken line in the figure) had large peaks in the initial stage of startup, but in the startup HWC. The peak of the initial startup disappeared and the oxidant concentration was maintained at a very low level throughout most of the startup operation.

  Therefore, it can be seen that performing hydrogen injection at the start-up of the BWR power plant has a great effect on reducing the occurrence of SCC.

  FIG. 7 is a diagram showing the corrosion potential (ECP) in the reactor water in the reactor pressure vessel when the BWR power plant is started.

The vertical axis indicates the temperature of the reactor water (° C.), the amount of hydrogen injected into the reactor water (Nm ³ / hr), and the corrosion potential (ECP) (V vs SHE). The horizontal axis indicates the elapsed time (hr) from the start of the start of the BWR power plant.

  The measurement position of the corrosion potential is the reactor pressure vessel bottom drain line 65. The corrosion potential is measured by using a corrosion potential sensor installed in the reactor pressure vessel bottom drain line 65.

  As shown in FIG. 7, the corrosion potential at the start-up HWC (plotted by the black square ■ in the figure) is lower than that at the start NWC (plotted by the white square □ in the figure). Except for the −0.23 V vs SHE (280 ° C. conversion) (broken line in the figure) level, which is an index of SCC sensitivity, was satisfied throughout the startup operation.

  In addition, the corrosion potential when the reactor water temperature is between 150 ° C. and 200 ° C., which is considered to be a region with particularly high SCC sensitivity, is reduced in the startup HWC, and the hydrogen injection method of this embodiment has a great effect. I understand.

It is a figure which shows each system | strain of the BWR power plant of this embodiment. It is a figure which shows the comparison result of the hydrogen injection position of the BWR power plant of this embodiment. It is a figure which shows the temperature change of the reactor water of the BWR power plant of this embodiment, and the change of the hydrogen injection amount. It is a figure which shows the SCC suppression effect of the BWR power plant of this embodiment. It is a figure which shows the chromate ion density | concentration in the reactor water of the BWR power generation plant of this embodiment. It is a figure which shows the oxidizing agent density | concentration in the reactor water of the BWR power plant of this embodiment. It is a figure which shows the corrosion potential in the reactor water of the BWR power plant of this embodiment. It is a figure which shows the flow volume of the reactor water of the feed water system at the time of starting of a BWR power plant.

Explanation of symbols

1: Reactor pressure vessel 2: Turbine 3: Condenser 4: Core 5: Hydrogen injection system 7: Reactor internal structural material 51: Recirculation system 54: Purification system

Claims (6)

Boiling water nuclear power that suppresses stress corrosion cracking of structural materials in the reactor in contact with the reactor water of a boiling water nuclear power plant that reaches the rated output by repeating the boiling and non-boiling states of the reactor water at startup A hydrogen injection method for a power plant,
When starting up the boiling water nuclear power plant, hydrogen is injected into the reactor water, and the amount of hydrogen injected into the reactor water when the reactor water is in a boiling state is determined when the reactor water is in a non-boiling state. A method for injecting hydrogen into a boiling water nuclear power plant , wherein the amount of oxidant in the reactor water is decreased by increasing the amount of hydrogen injected into the reactor water. Injecting the hydrogen from a system with continuous circulation of reactor water at the start-up of the boiling water nuclear power plant,
The hydrogen injection method for a boiling water nuclear power plant according to claim 1.   The method for hydrogen injection in a boiling water nuclear power plant according to claim 2, wherein the system having continuous circulation of reactor water at the time of startup is a purification system or a recirculation system.   The hydrogen injection method for a boiling water nuclear power plant according to any one of claims 1 to 3, wherein the hydrogen is injected so that a dissolved hydrogen concentration in the reactor water is 50 ppb or more.   The hydrogen injection method for a boiling water nuclear power plant according to any one of claims 1 to 3, wherein the hydrogen is injected so that a dissolved hydrogen concentration in the reactor water is 50 to 110 ppb. The method for hydrogen injection in a boiling water nuclear power plant according to claim 1 , wherein the amount of hydrogen injected in the boiling state is increased as the temperature of the reactor water increases.

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