JPS6337358B2 - - Google Patents

Description

[Detailed Description of the Invention] [Technical Field] The present invention relates to a method for measuring the local neutron flux in a nuclear reactor core housing elongated nuclear fuel assemblies arranged in a certain array pattern. The present invention relates to a method for arranging a detector and limiting local power density in response to a detection signal from the detector.

[Prior art and its problems]

In large nuclear reactors, detecting the local power (energy) density of the core (i.e., at each location) to ensure that operating conditions within the permissible limits, i.e., the limited power is not exceeded, at any location within the core. It is also necessary to limit the maximum power density at each location by an appropriate method. In addition, the reactor must be operated within predetermined design limits that, in conjunction with the power density distribution within the core, ensure sufficient margin to prevent core damage under both steady-state and transient conditions. It's getting old.

Fluctuations in the power density distribution that normally occur over a wide area within the reactor core can lead to mechanical distortion in the axial and radial directions of the core, so measurements that can detect and limit these fluctuations are necessary. The technology typically utilizes neutron flux density detectors inside and outside the core and core coolant temperature detectors.

Neutron flux detection technology plays an important role in detecting and controlling core power because there is a close correlation between neutron flux distribution in the reactor core and thermal power distribution. Therefore, many nuclear reactor designs utilize in-core neutron flux measurement systems to provide safety-related information in conjunction with power distribution. These in-core measurement systems typically fall into two categories: systems that use a large number of fixed detectors arranged in a uniform distribution throughout the core volume to provide data at individual points within the core; and systems that use neutron flux detectors that can be moved along a fixed path within the reactor core. In the case of fixed in-core detectors, the detectors must be extremely reliable because maintenance or replacement of the reactor requires shutting down the reactor. A type of detector known as a self-actuated neutron detector, which uses the radioactive decay process of a material that can be activated with neutrons to generate a detected output signal, is often preferred for use in a reactor core. A typical self-actuated neutron detector includes four parts: a neutron-sensitive emitter, an insulator, a conductive wire, and a sheath or collector. An emitter made of a material with a high thermal neutron active cross section, such as rhodium, is connected to an axially extending conductor. Both the conductor and the outer sheath are separated by a densely packed ceramic insulator that maintains high electrical resistance under reactor operating conditions. The conductor and outer sheath are designed to emit fewer beta particles than the emitter. The detector is grounded via an ammeter to allow measurement of a current proportional to the total beta radiation escape from the emitter when the detector is exposed to a neutron flux.

U.S. Patent Nos. 3,603,793, 3,375,370 and
No. 4,087,693 discloses various self-actuating neutron detectors, while U.S. Pat. No. 3,892,969 discloses a unique lead wire for use with such detectors. For example, the maximum ambient pressure is above 2300 psi (162 kg/cm ² ) in a pressurized small reactor design.
Below are the high temperature gas-cooled reactor designs.
It can range up to 700psi (49.2Kg/cm ² ). The ambient temperature condition in pressurized water reactor design is 650〓
(344℃).

The fixed self-operating in-core neutron flux detector is
It can be positioned at various locations within the core to measure the axial and radial neutron flux distribution to provide a history of power distribution and power fluctuations. Furthermore, these local neutron flux detectors can be uniformly spaced apart in the axial direction and configured as a detector assembly for measuring axial neutron flux gradients. These detector assemblies can then be positioned at various preselected radial positions in the core. Each individual detector thus measures the neutron flux in its vicinity.
During reactor operation, incoming neutrons pass through the outer shell and insulators and are absorbed by the emitters. Considering the case of an emitter consisting essentially of rhodium, the following reaction takes place: The emission of β particles from rhodium emitters and their migration is the source of the local neutron flux signal. rhodium
The capture of neutrons by 103 produces a radioactive isotope, rhodium-104, which has an unstable nucleus. After a period of time when the forces in the nucleus try to reach equilibrium, β
The particles are released and a daughter nuclide, palladium, is created. Palladium requires one more orbital electron than its parent to be balanced in charge. An electrical conductor (lead wire) in which the electron supply path for rhodium is grounded through the signal measurement circuit.
If so, the β decay rates (these are related to charge depletion) can be measured. The β decay rate is related to the neutron flux absorbed by the emitter. Most of the electrons escaping from the emitter penetrate into the insulator and end up staying in or around the outer sheath.
β-decay in the emitter brings a positive charge to the lead wire. The magnitude of the positive charge on the lead wire (which is directly related to the neutron flux absorbed by the emitter) is measured. The measured signal is not affected by electrons escaping from the outer sheath since the outer sheath and its surroundings are at ground potential. The useful life of the detector is primarily a function of the rate of rhodium depletion. However, this consumption rate is directly proportional to the consumption current and can be easily checked and replenished by known techniques.

The energy density distribution fluctuates throughout the core. It is therefore desirable to minimize fluctuations in the energy density distribution by appropriate feedback control while ensuring that operating limits or limiting outputs that are designed to provide sufficient safety margins are not exceeded. Ru.

The limiting power is the allowable operating margin, for example in relation to the departure from nucleate boiling (DNB) or the melting of the central pellet or the operating range of the emergency core cooling system (ECCS) or a combination of these. Defined as the marginal output. In other words, a limited output is an output above which it is still safe to exceed, but which requires the output to be reduced below. The control power can vary depending on location within the core and is selected through prior measurements.

Periodical magazine "KERNTFCHNIK" - 1974, Volume 16, 429-436
The paper ``Radioactivity Instrumentation for 1300-MW Pressurized Water Reactors'' published on page 1 discloses techniques related to detecting or limiting power density in nuclear reactors, in which signals related to the power density distribution within the core are detected. is obtained by combining and processing the signal from the out-of-core neutron flux detector and the signal from the reactor coolant thermometer. Disadvantages in this process arise from the use of out-of-core neutron flux detectors and the use of reactor coolant thermometers. This is because detailed model calculations must be coupled to relate these global parameters to individual local power densities within the core. This results in reduced detection sensitivity and relatively low detection accuracy. Furthermore, small variations in power density cannot be detected. That is, small variations are not compensated for by the model calculation.

In general, the use of in-core information in relation to safety concerns core offset (the difference in average core power between the upper and lower halves of the core) and the differences between each of the four parts of the core. It was limited to the generation of global parameters such as tailt (the difference from the average power between the four core sections in the axial or radial direction). These global parameters are determined by the DNB ratio (DNB
the ratio of the heat flux yielding to the maximum allowable heat flux) or
It is associated with limiting conditions such as ECCS heat rate (KW per fuel degree) and is then used to control said parameters. Various mathematical functions have been devised and known for determining the DNB ratio (DNBR). Because of the error in the solution resulting from global parameter-based control, it is common to be too conservative or take too much safety. A large body of literature related to the instruments, parameters and techniques used to detect or limit reactor power can be found in the Nuclean Power Reactor Instrumentation Handbook.
Systems Handbook, Volume 1 (1973) and Volume 2 (1974).

Methods and apparatus for handling nuclear reactor operational constraints are disclosed in US Pat.

As such, the reactor core and associated coolant, control and protection systems are protected from any conditions of normal operation, including the occurrence of various operational events that are expected to exceed specified power limits. The design is designed with a reasonable margin to ensure that this limit is not exceeded even during periods of time. If a limit set point for control or monitoring is specified for a variable condition (a condition that involves control of output), the most severe abnormal situation that can be expected must be corrected before the limit is exceeded. Adjustable set points and adjustment techniques must be established to prevent the occurrence of.

[Purpose of the invention]

An object of the present invention is to provide a power limit control method for operating a nuclear reactor without excessive safety.

[Summary of the invention]

The present invention establishes a distribution of limit outputs for each zone, which represents the operating output at the permissible limit, and calculates the difference or product between the distribution of limit outputs and the sensed local output within the zone. determining a tolerance that is a comparative value of the core, and establishing a set point for the in-core instruments using the minimum tolerance that is the smallest of the tolerances and the local power representative of the current state of the area; It is aimed at internal power limit control technology, which means that even if the local power density or power distribution increases beyond the limited power in a particular core region, only that region needs to be returned to within the limited power. , which offers the benefit of not requiring a complete shutdown of the furnace.

In one embodiment, the invention includes generating a localized output signal (representative of a localized core area) derived from measurements of an in-core neutron flux detector. This signal is transformed to simulate the maximum local output signal for the most severe of the local output signals in the area. For example, choose the maximum value in the local output signal within the area. then a limit power value (a pre-established maximum allowable power value for said predetermined range of areas, which typically includes an area covering a portion of several fuel element assemblies, above which a maximum allowable power value should be returned within that range); A tolerance is determined based on the difference or ratio between the converted maximum local output signal value and the converted maximum local output signal value, and then a minimum tolerance, which is the minimum value of the tolerance in each area, is determined. after that,
The set point associated with the sensor is generated as the difference or product of the local output signal from the detector in that area and the minimum tolerance, or as a function of the local output and the minimum tolerance.

In another embodiment, the invention includes generating a localized output signal derived from measurements of an in-core neutron flux detector. This local output signal is compared to the limiting output signal in the area. The tolerance is determined based on the difference or ratio between the local output signal and the smallest of the limiting output signals in this area, and then the smallest tolerance therein is determined. Thereafter, the set point associated with the sensor is determined as the product or difference of the local output signal of the sensor in its covered area and said minimum margin, or as a function of the combination of the minimum tolerance and the local output signal of the sensor. caused to occur.

In conjunction with each of the above embodiments of the technology of the present invention,
If the parameter becomes greater than the set point, a signal initiates an appropriate action, such as power reduction by control rod movement or reactor shutdown by control rod scram. According to this method, even if the local power exceeds the limit power, the set point of the in-core instruments can be continuously corrected as long as the increase in the local power and the overall power at that point is safely acceptable. .

One of the objects of the present invention is to provide a method for improving power density detection sensitivity and detection accuracy, which has the advantage of being able to deal with variations in the power density distribution that have not been determined in advance, and which can be tolerated. To provide a method for preventing the occurrence of invalid operating values.

The method of the invention can be used to automatically reduce reactor power in conjunction with a control system. The method of the present invention locates or limits local power density to optimize power distribution within the core.

Another benefit of the method is that the power density within the core is limited to below a local maximum allowable value, regardless of where in the core the maximum allowable power density value is achieved.

In the method of the invention, the area detected by the individual detectors preferably covers a plurality of fuel element assemblies in the radial core plane, up to one quarter of the fuel element assemblies in the core. These probed areas may be overlapped to some extent to provide multiple extra constraints on the maximum local power density within the core. In a particularly elegant technique, the detectors are axially arranged so that the maximum local power density is detected as a direct function of axial height in the core.

[Explanation of Examples]

A detailed explanation will be given below with reference to the drawings.

Referring to the drawings, FIG. 1 illustrates a cross-section of a typical pressurized water reactor, in which a sensor for neutron flux in the reactor core is associated to automatically adjust set point values. The in-core power limitation control system and technique of the present invention can also be used to limit or detect in-core power. For convenience of explanation, the present invention will be described herein in connection with a pressurized water nuclear reactor; however, the present technology is not intended to be limited to any particular type of reactor design.

The nuclear reactor of FIG. 1 includes a pressure vessel 20 containing and supporting a reactor core 21. The reactor core 21 is
It consists of an arrangement of mechanically interchangeable fuel element assemblies 22. The fuel element assembly 22 includes a cylindrical core support structure 2, as shown in FIG.
They are arranged symmetrically within 3. The support structure 23 includes the fuel element assembly 22 and the lower grid assembly 2 on which it is placed.
4 and a downward flow distributor 25. An inlet nozzle 30 and an outlet nozzle 31 provide means for the introduction and removal of coolant into and out of the reactor pressure vessel 20, respectively. Inlet nozzle 30
The reactor coolant entering the pressure vessel via the pressure vessel flows downwardly through an annular region formed between the pressure vessel interior surface and the exterior surface of the support structure 23. The flow is carried out by the distributor 25
is distributed into the core 21 through. The coolant flows along the fuel element assembly 22 and removes heat therefrom, and the heated coolant ends up at the outlet nozzle 3.
1 and then released from the pressure vessel 20.

Each of the fuel element assemblies 22 includes a lower end fitting 34 and an upper end fitting 35, a plurality of elongated fuel elements 40, a hollow cylindrical guide tube 41, and a hollow cylindrical guide tube 41, as shown in FIGS. A central instrumentation tube 42 is included. The fuel elements 40, guide tubes 41 and central instrumentation tube 42 are arranged in a rectangular parallel arrangement by one or more spacer assemblies 43 spaced along the length of the fuel element assembly. Supported at intervals in the transverse direction. The ends of guide tube 41 are mechanically secured to lower and upper end fittings 34 and 35.

A plurality of nozzles 44 are positioned within the lower head of pressure vessel 20. Each nozzle 44 is connected to a tubular in-core instrument guide member 45, the latter eventually being operatively connected to one of the fuel element assembly instrumentation tubes 42.

As best seen in FIG. 2, assemblies 46 of self-actuated neutron flux detectors 50 are uniformly distributed at preselected locations within core 21. As shown in FIG. Each assembly 46 includes a plurality of neutron flux detectors 50 distributed at uniform axial spacing (as shown in FIG. 1) through the reactor core 21 . Each detector assembly 46 is positioned within the central instrumentation tube 42 of the fuel element assembly 22. As discussed below, at least some of the in-core neutron detectors 50 rely on the positioning of control rods within the guide tubes 41 to compensate for changes in reactivity due to fluctuations in operating conditions and to shut down the reactor. provides a signal that can be used to adjust the Additionally, these neutron flux signals may be modified to eliminate or compensate for background currents by means and techniques well known in the art. Although the inventive technique described below does not require special positioning of the detectors 50, the detectors 50 should be uniformly distributed throughout the core 21. Each detector provides a measurement signal that is proportional to the neutron flux in its vicinity and is used to determine the local power density.

According to a preferred embodiment of the invention, nucleate boiling and
ECCS linear heat rate (per unit length of fuel rod)
A limiting power distribution, such as a permissible deviation limit from KW), is determined throughout the core. The limiting power distribution is selected to provide sufficient safety to ensure that critical operating conditions are not achieved. Therefore,
The local DNB ratio (DNBR), defined as the ratio of the heat flux resulting in a deviation from nucleate boiling (corresponding to the limiting power) and the actual heat flux at a particular core location, is:
It is usually limited to a minimum value (for power generating reactors) during steady state operation, normal operating transients and expected transients. Similarly, the ECCS heat rate (KW/m) limit is based on a combination of radial and axial limiting power that prevents melting of the central fuel. Although DNB is not a measurable parameter, the measurable parameters of neutron energy, reactor coolant flow, temperature and pressure can be related to DNB through suitable known correlations, both axially uniform and non-uniform. Determine the position of DNBR and DNB relative to the heat flux distribution. Signals relating to local power density are obtained from electrical signals generated by at least some of the assemblies of in-core neutron flux detectors 50 located in the central instrumentation tube of each fuel element assembly 22. These sensor assemblies form part of the In-Core Limit Control System (ILCS). As schematically illustrated in FIGS. 5 and 6,
Each of the ILCS assemblies 46' detects each of several zones 51A-Q defined within the core, hereinafter referred to as distribution zones, containing multiple fuel element assemblies. The zones are selected to substantially uniformly divide the core when viewed from above, and a vertical row of sensors 50 is located approximately at the center of each zone. In the illustrated embodiment, the representative area 51 consists of seven axial sections of equal volume with different axial levels L1~
7 and one sensor 5 for each section L.
0' is equipped. As illustrated in FIG. 7, the local power density or detector output 60 is determined as a function of the sensed neutron flux by the sensor 50' at each axial level L.

The output signal provided by the ILCS sensor as well as
A signal relating to the power distribution in substantially all of the fuel element assemblies for the ILCS measurement area is preferably transmitted to an on-line computer. The sensed ILCS output heat rate or local power signal is then compared to the limit power 61 to determine the tolerance (difference between the local limit power and the sensed local power in the axial segmented volume) in each zone. Determine. Select the minimum tolerance for all fuel element assemblies in each axial section that is the smallest of the ratios of the limiting power conditions (such as DNBR or ECCS heat rate tolerance limits) to the sensed local power density. It is determined by Alternatively, as illustrated in FIG.
The minimum tolerance curve 62 may be characterized as a function of the maximum relative allowable increment in heat rate (until an operating limit is achieved) versus zone height.
The minimum tolerance is used to adjust the setpoint in conjunction with each ILCS sensor signal, e.g., to initiate a power reduction (through adjustment of control rod position) when the signal exceeds the setpoint and when the signal exceeds the setpoint. Used to stop output reduction when the value is below the value. Thus, as shown in FIG.
7 DNBR and associated with each division of 6'
ECCS heat rate set points exist. If the segmented output at a certain level exceeds the set point, a breached tolerance condition occurs. Many reactor parameters depend on more than one variable and
It should be appreciated that set points, such as for DNBR limits, are continuously adjusted or updated.

An on-line computer is used to continuously detect and adjust the set point and its tolerances based on signals received from the in-core sensors. The set point can be set based on the DNBR limit or the ECCS linear heat rate limit (limited output) or both. If the set point is exceeded, appropriate action is initiated to reduce the output.

In a preferred embodiment, data in the form of electrical signals needed to determine the DNBR and ECCS heat rate set points is communicated to a set point generator module. The set point generator module calculates the DNBR and ECCS heat rate limit power (Kw/ft or Kw/m) as a function of axial height for each fuel element assembly in the core. The measured operating heat rate (obtained from the ILCS sensor) is compared to these limit values by on-line calculation to determine each level (category) in each division.
Determine the margin difference in . The minimum margin difference may then be determined by selecting the minimum value of the ratio of the constraint to the operating condition at each level for each distribution area. This defines the relative allowable increment in the axial level (segment) output of the detector.

The DNBR or ECCS heat rate set point is then calculated as the product of the current detector output value and the minimum tolerance.

On the other hand, in the present invention, DNBR or ECCS
The heat rate set point can also be calculated by adding the current detector output to the minimum tolerance. The minimum tolerance in this case is equal to the minimum difference between the limit power and the operating power.

In other words, in Figures 7 and 8, the minimum tolerance is calculated as either the minimum value of the limit power 61 divided by the sensor section power 60 corresponding to the local power density, or the minimum value of the difference between the two. It can be done.

The detector measurement signal, together with associated limit values, can be analyzed by analogue or digital means or both in the following manner: First, the detector measurement value in the immediate vicinity of the detector is converted into an output. These powers are converted to maximum fuel rod powers in a given band under observation (preferably covering multiple fuel element assemblies and up to 1/4 of the core). The maximum local power thus determined is compared to the limiting power for the local band (mostly a function of axial height) to generate a margin and adjust the set point at the location associated with the detector.

If self-actuated in-core neutron flux detectors equipped with rhodium emitters are used, for example to prevent transient reactions after the start of energy reduction by the in-core power limit control system, the individual detection signals S
is modified as a function of the current detector output Pt and the detector output Pt′ at a certain point after the start of the reactor power reduction, and the reactor power reduction is stopped when SPt′/Pt<set point. . The power level at which the detector signal S is stable is Pt = Pt'(1-e ^- 〓 ^t ), where λ is the value reflecting the 43 second half-life of the power decrease and t is the value from the start of the power decrease. (representing the time until Pt′ occurs). The detector signal is said to be "stable" when it is reduced to a value less than the set point. Once the set point is exceeded, power reduction is initiated and power reduction is stopped when the detector signal is again less than the set point, ie, becomes stable.

Although the present technique has been described in connection with adjusting set points for the purpose of limiting power density, it is also possible that such set points are used to indicate that a parameter or condition under monitoring has reached a selected value. The methods described herein may be used to detect local power densities, for example by issuing alarms, or to provide an indication of the state of local power densities or distributions to provide sufficient information for operator control. It will be apparent to those skilled in the art that the set points may be similarly adjusted.

[Brief explanation of drawings]

Figure 1 is a schematic vertical sectional view of the reactor, showing the relative positioning of the fuel assembly and neutron flux sensor, and Figure 2 is an enlarged horizontal sectional view taken along line 2-2 in Figure 1. 3 illustrates a core containing 205 fuel element assemblies arranged symmetrically in four quadrants of a cylindrical member; FIG. 3 shows a fuel element assembly of the type used in the reactor of FIG. FIG. 4 is a schematic cross-sectional view of the fuel element assembly of the type illustrated in FIG. indicates the location,
FIG. 5 is a schematic plan view of a core containing 205 fuel element assemblies illustrating the division of sensor assemblies used in accordance with the present invention, and FIG. FIG. 7 is a graph representing the per unit axial length limit and operating power density for the one distribution area, and FIG. FIG. 8 is a graph representing the relative allowable increments in heat rate from the operating conditions of FIG. 20: Pressure vessel, 21: Core, 23: Core support structure, 30, 31: Inlet and output nozzle, 2
2: fuel element assembly, 40: fuel element, 41: guide tube, 42: instrumentation tube, 44: nozzle, 45: in-core instrument guide member, 50, 50': neutron flux sensor,
46, 46': Sensor assembly, 51A-H: Sharing area, L1-7: Axial section, 60: Sensor output curve, 61: Limiting output condition curve, 62: Minimum tolerance curve.

Claims (1)

[Claims] 1. For detecting or limiting local power density in a nuclear reactor core equipped with neutron flux detectors with uniform distribution throughout the core containing a plurality of fuel element assemblies. A method for adjusting the set point values used, the method comprising: pre-establishing a limiting power signal representative of operating conditions at acceptable limits within each preselected region of the core; generating a local output signal by each of the internal neutron flux detectors;
Compare the local output signal and the limit output signal and determine the tolerance between them, select the minimum tolerance within each zone, and set the value of the set point to the minimum tolerance and the local output signal. A method for adjusting the in-core power limit as a function of. 2. The method of claim 1, wherein the margin difference is determined by comparing the most severe local output signal in the preselected area with the control output signal. 3. The method of claim 2, wherein the preselection area includes at least a common portion of a plurality of fuel element assemblies. 4. A method according to any one of claims 1 to 3, wherein the margin difference is determined as a function of the difference between one of the limiting output signals and the local output signal. 5. The method of claim 4, wherein the set point is adjusted as a function of the minimum tolerance and the sum of the local output signal. 6. The method according to claim 5, wherein the limiting output signal is a measure of deviation from nucleate boiling. 7. The method of claim 5, wherein the limiting output signal is a measure of the linear heat rate of the fuel element. 8. Claim 1 in which the tolerance is determined as a function of the ratio of one of the limiting output signals and the local output signal.
The method described in any one of 3 to 3. 9. The method of claim 8, wherein the set point is adjusted as a function of the product of the minimum tolerance and the local output signal. 10. The method of claim 9, wherein the limiting output signal is a measure of deviation from nucleate boiling. 11. The method of claim 9, wherein the limiting output signal is a measure of the linear heat rate of the fuel element.