KR830001313B1 - Protection of Pressurized Water Reactor - Google Patents

Abstract

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Description

Protection of Pressurized Water Reactor

FIG. 1 shows the reactor power level with respect to the nominal power as a function of time during three days of operation of a pressurized water reactor in the case of a low plato with an output equal to 90% of the nominal power.

FIG. 2 shows a change in the maximum value of the output per unit length along the fuel element portion during the operation period corresponding to the graph as in FIG.

3 shows the corresponding change from the margin relative to the starting point of protection against boiling.

4 shows the change in output measured for nominal power during a three-day operating cycle with a low plato at a value equal to 50% of nominal power.

5 shows a corresponding change in maximum power per unit length according to fuel element portion.

FIG. 6 shows the corresponding change in margin with respect to the onset of protection against boiling.

7 shows the development of the axial deviation of the flux during the instantaneous return to the nominal output after several cycles of operation at 50% of the nominal output when operating the reactor between the lower plateau and 50% of the nominal output. The figure which shows.

FIG. 8 shows the development of maximum power per unit length along the fuel element portion during operation between nominal power and plato at 50% of nominal power as an instantaneous return of power every hour.

9 shows the corresponding development of a margin relative to the point of initiation of protection against a crisis.

The present invention relates to a method for protecting seizure reactors that avoids exceeding operating limits such that the entire fuel is not disturbed by the occurrence of boiling crisis or melting of one or more fuel pellets.

In a pressurized water reactor, the fuel element portions are arranged parallel to each other in the core of the reactor, and control rods of material that absorb neutrons for control of the reactor move between the fuel element portions. The fuel element is submerged in the primary fluid, which is water under pressure, which serves as heat transfer and moderator.

During the operation of the reactor, if the amount of heat is too large to be absorbed and not carried by the heat carrier fluid or the flow of the heat carrier fluid is insufficient, it may interfere with the entire fuel element section performing irreversible drops under the thermal effects they emit. It is necessary to avoid things.

In the control of pressurized water reactors, people have sought to prepare for two other phenomena that involve some complete destruction of fuel. Two phenomena are, on the one hand, the boiling crisis of water under pressure in contact with the fuel element portion, and on the other hand the melting of the fuel.

As a result, when the heat exchange between the water and the fuel under pressure across the outer surface of the fuel rod, the high temperature of the fuel element portion in the near boiling state despite the water is maintained at a high pressure. Limited boiling near the fuel element always permits heat exchange between the fuel element and the water under pressure. However, it is necessary to avoid such boiling, which forms a film of steam along the fuel element section, because a single film of steam prevents large exchanges between the fuel element section and water under pressure by the phenomenon of nominal conditions.

Therefore, it is very important to watch the onset of boiling or the appearance of boiling crisis by the generation of steam in contact with the fuel element portion, in order to avoid the occurrence of an operating zone in which the boiling crisis occurs while driving the reactor.

Indeed, until now, in order to consider the method of evaluating the measurement conditions and the accuracy of the boiling crisis, efforts have been made to place a margin of safety between the operating conditions of the reactor and the conditions in which the boiling crisis occurs.

On the other hand, fuels are caused by too large emissions of power, which can occur at the level of the fuel elements and cause their destruction to occur as a local rise in temperature reaching the melting point of the material forming the fuel pellets. One or more pellets of the urea part are melted. Melt is induced when the release of this output exceeds a value per unit length or a certain starting point, determined by the fuel's characteristic function.

The protection technology of pressurized water reactors for the emergence of either of these two phenomena has thus far consisted of calculating the variables indicative of these phenomena and initiating an emergency stop when one of the variables exceeds a specified limit, which is measured. Consider any inaccuracies, representative characteristics of the variables, and the accuracy of the measurement of these variables.

Thus, if the determination of the variable is done with some inaccuracy, it may be possible to initiate an emergency stop when the actual threshold has not been reached.

Variables and linear outputs representing the boiling crisis are determined by calculation from the measurement of the inlet and outlet temperatures of the coolant of the reactor, from the measurement of the pressure and flow rate of this coolant and the distribution of the core output of the reactor.

In order to increase the accuracy of the determination, it is apparent that the accuracy of the measurement of temperature, pressure and flow rate is increased or until now the distribution of output in the core of the reactor is not determined with accuracy during the operation of the reactor, and whether it is an axial distribution or a radial distribution. I was not satisfied with using the superficial values for this distribution.

As a commonly used method for handling pressurized water reactors, attempts have been made to keep the flux axial deviation at a constant value.

_H와 노심의 하부로 부터의 중성자 플럭스

_B사이의 불평형을 나타내는 변수이다. 이러한 변수는 비율 t-1 과 같다.The axial deviation of the flux is the neutron flux from the top of the core

Neutron Flux from _H and Bottom of Core

Variable representing the unbalance between _B This variable is equal to the ratio t-1.

Thus in this method of handling at constant on-axis deviation, as long as the distribution of flux or output in the core is concerned, only one measured variable is the on-axis deviation of the flux.

However, for one and the same value of the axial deviation, one can simulate the shape of the flux distribution along the extremely variable fuel component.

To account for this uncertainty about the distribution, an extremely poor distribution is used for the calculation of the variables for boiling and linear output and is called the superficial distribution, which can have all of the best cases of the distribution, one of the fluxes. And induces the same axial deviation.

As far as the radial distribution of the flux is concerned, all possible states are taken into account and the apparent factor of the distribution as a function of the position of the control rod.

Therefore, even if such a superficial distribution is not necessarily a critical case, it leads to a relative disadvantageous evaluation of a variable that can lead to an emergency stop.

On the other hand, in the handling of other reactors from a constant axial deviation, the handling in a constant axial deviation with the absorption control group, as described in Applicant's French Patent Application No. 77-19316, which has been conveniently used so far. In the case of using a group of control rods that absorb less than the group employed, it is no longer possible to use a apparent value that takes into account the position of the control rod for the radioactive distribution, which is the position and output of the control group. This is because both positions of the absorption group for recovery are taken into account, which is either extremely complex or only the position of the output recovery group, which leads to a very approximation of the variables. In this way, not all are benefited by the flexibility of the use of the reactor connected to this handling.

It is therefore an object of the present invention to provide a method of protecting a pressurized water reactor having fuel element portions disposed parallel to each other between control rods moved for control of the reactor and inside the core of the reactor, which method is characterized by It is possible to avoid exceeding the limits of the operation in which the whole fuel is impeded by the appearance of or the melting of one or more fuel pellets, and a variable indicative of the output radiated per unit length of the boiling element or fuel element along the fuel element is conveyed. When the limit is exceeded, an emergency stop is initiated and the parameters are determined from the measurement of the inlet and outlet temperatures of the coolant in the reactor, the measurement of the pressure and flow rate of such coolant, and the output distribution of the reactor core, which is more appropriate. Emergency stop can be initiated in the form and this can be A handling method of the reactor is fully utilized to enable flexible control of the character.

For this purpose, the output distribution in the reactor core is determined by measurement to approximate the actual physical limits of the use of the reactor. In other words, the ion box disposed outside the reactor core includes measurement of the axial distribution of the output, measurement of the output distribution in the longitudinal direction of the fuel element portion, measurement of the position of the control rod that determines the radial distribution of the output, and the like.

By way of example and not by way of limitation, embodiments are described, where the measurement is carried out to enable the protection method according to the invention to calculate the variables representing the output and the boiling per unit length, and the axial distribution of the output and the radial distribution of the output determine the position of the control rod. Marking and affecting the pressurized water reactor, which is determined by taking account of the measuring room.

Another method of work described later is to carry out a device on a pressurized water reactor for the measurement of neutron fluxes which consist of ion boxes which are arranged axially one above the total height of the core and on top of the other outside the reactor. do.

Such a measuring device with an ion box has already been described in French Patent No. 2,268,353 and has six measuring chambers with two electrodes, one of which is covered with a precipitate of boron.

Each of the six measuring chambers is in contact with an electrode maintained at a different voltage and surrounds the ionizing gas.

)에 의해서,

입자가 생성되고, 발생된 이온이 전극을 향하여 구동되고 모아지는 결과로 실내에 모아진 가스를 이온화시키면, 이러한 전극의 높이에서 방출되는 전류의 측정이 이온상자의 높이에서 원자로에 의해 방출되는 중성자플럭스의 측정을 나타낸다.Upon impact by neutrons traveling from the core of the reactor, upon precipitation of boron carried by one of the electrodes (n,

),

When particles are generated, and the generated ions are driven and collected toward the electrodes, ionizing the gas collected in the room, the measurement of the current emitted at the height of these electrodes results in the neutron flux emitted by the reactor at the height of the ion box. Indicates measurement.

From the ionization current measured at each outdoor height, the neutron flux is determined at the height of the precipitate of boron which is carried out in the measurement stall.

So far, for the purpose of measuring the neutron flux in wells whose position is defined along the height of the core, the precipitation of boron has been carried out in the measuring chamber and is of short length in the axial direction. This was handled with loss of accuracy and sensitivity of the measurement and was biased in obtaining an accurate image of the axial distribution of the output in the core.

In the case of a measuring device having an ion box used in the scope of the present invention, the deposition of boron was carried out into each box over the entire height of such a box with a gap between the boron deposits in each box of short length, The total length of the boron deposits in the six ion boxes became close to the length of the fuel element.

In practice, six measuring sections 46.4 cm long with 4 cm spacing were used.

From the measurement of the ionization current in each box, a new calculation along the fuel element is used, using the expansion of the polynomial function representing the axial distribution of the output, ie the change in output that is a function of the axial position of the developed point. This allows you to get an accurate representation of the axial distribution of the output.

In addition to these measurements that allow the determination of the axial distribution of the outputs, the control rods are positioned like the measurement of variables that allow accurate measurements to determine the output of the reactor.

Variables that can determine the output value are the flow rate of the fluid measured for the pressure of the primary fluid, the inlet and outlet temperatures of the fluid in the reactor, and the nominal flow rate. From these variables, the position of the control rod group in the direction of the core height is determined and the output value of the reactor, which allows the radial distribution of the output to be determined according to the position of the control rod in the direction of the core height, is determined in the conventional form.

From the output of the reactor and the axial and radial distributions of the output, one can determine the output per unit length along the fuel element for each axial region defined by the height of the core.

An emergency stop is initiated if one of the values of the output per unit length exceeds the limit output determined as a function of the fuel's characteristic.

H의 증가를 위한 계수를 결정할 수 있다.From the axial and radial distributions of the output, enthalpy F

The coefficient for the increase of H can be determined.

H의 증가를 위한 계수로부터, 원자로 출력값 Q로부터, 그리고 이온상자속에서 수행되는 측정으로부터 결정 한계를 결정할 수 있으며, 비등위기에 근사하는 경우에 비상정지를 결정할 수 있다.Enthalpy F

From the coefficients for the increase of H, from the reactor output Q, and from the measurements performed in the ion box, the decision limit can be determined, and the emergency stop can be determined in the case of approximation to the boiling crisis.

This calculation is performed for a channel that represents only the core of the reactor, which channel is called a hot channel. This makes it possible to simplify the resin processing of data even if all of the output distributions are known in the core and more complicated processing remains.

Thus, the method according to the present invention can determine an emergency stop if a certain variable exceeds the value exceeded by them, and this variable is determined by accurately considering the determination of the output distribution in the core without the use of the apparent distribution, Can protect them.

Thus, in any operating state of the reactor, it is possible to determine the output per unit length of the fuel element as a function of the core position and the maximum value of this variable during any operation of the reactor.

Thus, in the case where the reactor corresponds to the first degree of operation at an output equal to 90% of the nominal output or at nominal output, it is possible to determine the operating state, the maximum power per unit length, the change indicated in the second degree, and so forth.

It can be seen that the operation of the reactor corresponding to the graph shown in FIG. 1 goes to the output per unit length and does not exceed the limit indicated by the horizontal line of the longitudinal axis such as this limit (425 W / cm).

FIG. 3 shows the margin for the starting point where the change initiates protection against boiling crisis during the operation of the reactor according to the output program shown in FIG. Similarly, it is shown that the margin remains sufficient to operate the reactor which can be followed for three days corresponding to the output program.

Figure 4 shows a similar output program of three days during the course in which the reactor is used at 50% of its nominal output, or at its nominal output, and has a momentary feedback to this output between the two operating states.

In FIG. 5, the corresponding change shows the maximum power per unit length in units of CW per foot and Watts per cm.

It can be seen that operation is possible so as not to reach the limit value at which an emergency stop is initiated.

6 shows the corresponding change in the margin relative to the starting point for initiating protection against the boiling crisis, and it is shown that it is possible for the operation not to reach the lowest limit of the limit of safety.

In FIG. 7, as a function of the variable representing the axial deviation, as a required output graph of the reactor against the nominal output, the method illustrates the operation of the reactor between nominal and 50% of the nominal output, and 50% of the nominal output. Maintaining the furnace reactor is placed for increased periods of 2, 6, 9, 10 and 12 hours.

In this group, the curve is shown as a trapezoidal solid line of remaining work, which illustrates the operation of the reactor to avoid the onset of emergency stop when a conventional method of protecting nuclear reactors is used.

After maintaining at 50% of nominal power for more than 9 hours, it can be seen that the reactor remains trapezoidal, making it impossible for the reactor to immediately return to its nominal power. Thus, the conventional method leads to an emergency stop exceeding the job limit value.

In FIG. 8, it is shown that the graph corresponds to the maximum output per unit length as a function of time for the operation of the reactor between 50% of its nominal output and the nominal output immediately returning to the nominal output at any moment. If the immediate return of is carried out every hour and maintained at 50% of the nominal output, it is raised to the increase in output per unit length represented by the point placed on a continuous curve indicating the development of the linear output.

Therefore, it can be seen that the limit value of the maximum output per unit length never exceeds when the feedback becomes the nominal output.

Similarly, FIG. 9 shows that when a corresponding development returns to nominal power at any moment, a margin is shown for the initiation of protection against boiling and this margin remains above the expected lower limit value. Lose.

The last embodiment shown in FIGS. 7,8 and 9 shows an increase in the operation of the reactor in view of the determination of the axial and radial distribution of the output, which allows the method according to the invention to allow more accurate calculations of the variables representing useful limit values. Demonstrates providing flexibility.

As a result of exceeding the operating limit value for one of the variables representing the outputs per boiling or unit length, an alarm is initiated and, depending on the other variable and its value exceeded, prevents the withdrawal of control rod bundles or leads to an emergency stop. Allow the turbine to reduce its load automatically.

Determination of the initiation of alarms with other variables is obtained with the consideration of the microprocessor which enables the progression of the other variables measured to infer the observed safety variables from them. The performance of the results allows for the observation of the development of other variables related to the protection of the reactor and the alarm due to exceeding limit values for the observation of the development of other variables.

For the purpose of facilitating the acquisition of variables, it can be seen that the calculation is performed on a channel representing the core of the reactor called a hot channel. However, if long latencies are readily available, it is also possible to use large computers to calculate variables for the whole core.

In other words, the present invention is not limited to the described embodiment but includes any change.

Therefore, it is also possible to use the polynomial form described for the determination of the axial distribution of the output from the measurement of the ion box current when the measuring chamber uses a region of sensitive material of high height, which gives an axial distribution representing the actual state of the core. Other forms of interpretation are also possible which are satisfactory to obtain and allow the calculation of this distribution.

As far as the calculation of the radial distribution of the output is concerned, the direction of its height from the position of the bundle of control rods to the determination of the presence or absence of each of the groups of control rods at the value of the part under consideration when the method calculates the radial distribution coefficient. A compartment of the core (eg, 30 compartments) is performed in the stomach. Other forms of indicating the position of the control rod under conditions that measure each phase of the reactor's operation are also understood.

In addition, the present invention is not limited to the use of a microprocessor for each determination of each intermediate variable and a variable indicative of a working limitation, and a larger capacity computer can be used to perform more complex variable determination.

Any possible form of alarm or emergency device initiated from the comparison of the assigned limit value and the variable may be used and inevitably initiates reverse automatic operation or manual control of the control element of the reactor.

Finally, the protection method according to the invention can be applied to pressurized water reactors in any handling method or in any required output program.

Claims (1)

The distribution of output in the reactor core, the flow rate and pressure of the coolant, the inlet and outlet temperatures of the coolant in the reactor, and the limits of the fuel fuel being impeded by the appearance of a boiling crisis or melting of one or more fuel pellets. A method configured to initiate an emergency stop when an output emitted per unit length of a fuel element determined from a measurement or a variable representing an boiling crisis along the fuel element exceeds a specified limit, so that the actual physical limit for use of the reactor is possible. In order to approach closely, the output distribution of the reactor core is measured by measuring the axial distribution of the output, that is, the longitudinal output distribution of the fuel element portion, by determining the radial distribution of the output and measuring the position of the control rod by an ion box installed outside the core. Method for protecting a pressurized water reactor, characterized in that determined.

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