Field of the Invention

The present invention relates to nuclear power generation and can be used in operation of nuclear power reactors with a liquid heat transfer medium.

To simplify the metering and transportation of nuclear fuel in a reactor, fuel cells containing nuclear fuel are assembled into fuel assemblies. When operating the reactor, the nuclear fuel re-loading procedures are of particular importance. The nuclear fuel reloading includes a series of procedures to remove and transport fuel assemblies in order to replace or rearrange them in the reactor. During re-loading stages, the fuel leakage detection of fuel cells collected in a fuel assembly is performed. Fuel leakage detection can be performed for all fuel assemblies or for particular fuel assemblies selectively.

Fuel cell tightness is necessary to preclude fission products from getting into the heat transfer medium, which may lead to radioactive elements spreading out of the core region. Furthermore, due to the fact that uranium, plutonium, and compositions thereof, being a part of the fuel cell core, are highly reactive, the chemical reaction thereof with water can lead to fuel cell deformation and other undersirable consequences. Fuel leakage detection serves to detect fuel assemblies with non-tight fuel cells that are subsequently excluded from service or repaired. Widely used methods for fuel leak detection of fuel cells in fuel assemblies, performed in FFDS canisters, are based on placing fuel assemblies into an enclosed volume filled with borated water, forcibly discharging fission products from non-tight fuel cells into said volume, and taking samples from said volume, thus measuring reference radionuclide leaks from non-tight fuel cells by e.g. measuring water samples.

Taking into account transport and technological procedures, the execution of conventional fuel leakage detection significantly increases duration of nuclear fuel reloading, and consequently, the reactor idle time, which is an obstacle that needs to be overcome in order to increase the total capacity efficiency of nuclear power plant, and also requires additional borated water consumption and increases the amount of liquid radioactive waste.

RU 2186429 discloses a method and system therefor used to decrease reactor idle time by simultaneously performing- reloading and fuel assembly leakage detection procedures. According to the method disclosed in RU 2186429, a volume of gas previously pumped into a vessel is passed through the heat transfer liquid, and the gas is sampled at a gas space point directly above the heat transfer medium surface, wherein the gas supply and sampling are performed both during elevating the fuel assembly towards the transport condition and during horizontal transport of the fuel assembly. The gas sample is supplied to the measuring device through a pipeline using a pump. The radioactivity measurement of the sample is performed continuously using a detector provided inside the measuring device; the detector signal is input into the data processing and display unit. The presence or absence of non-tight fuel cells is concluded based on the value of said signal. The disclosed solution allows to reduce duration of nuclear fuel reloading by combining the fuel cell leakage detection and fuel assembly transport, however, the fuel leakage detection results are of insufficient reliability. In particular, it is possible to miss, i.e. fail to detect a non-tight fuel assembly due to performing detection during the fuel assembly horizontal transportation procedure. Gas containing radionuclides released from damaged fuel cells moves up into the volume above the heat transfer liquid surface. Due to the volume in which the fuel assembly is placed being non- tight, there is a chance of gas or part thereof not passing into the above-liquid volume, from which the samples are being taken, during horizontal transport of the fuel assembly. Consequently, a sample thus taken can be practically indistinguishable from the sample corresponding to a tight fuel assembly. Furthermore, in the method of RU 2186429, the measured radioactivity is not divided into various types, such as β-activity and γ-activity, which can lead to false non-tight fuel assembly detection. Also, the known technical solution lacks sample representativity control and means therefor, which also reduces reliability of fuel assembly leakage control. Hereinafter, the term "sample representativity" refers to said sample meeting the criteria of reliability for measurement results within error margins for measuring equipment used.

The object of the present invention is to provide a method for fuel leakage detection for fuel cells of fuel assemblies that not only allows to decrease the reactor idle time by combining fuel assembly leakage detection and the reloading procedures, but also provides high reliability of fuel assembly leakage detection results.

The object is solved by a method for fuel leakage detection of the fuel assemblies during nuclear fuel reloading in a reactor with heat transfer liquid that comprises the following actions:

The overall background β- and γ-activity is determined prior to commencing reloading. To this end, at least two samples are taken at random points above the reactor and the fuel pool, and the average backround β- and γ-activity values are determined based on sample radioactivity measurements. The nuclear fuel reloading in a reactor with heat-transfer liquid includes successive reloadings of each fuel assembly subject to reloading. The reloading of one fuel assembly subsequently includes installing the fuel assembly transport device comprising an outer and at least one inner section in position to remove the fuel assembly, placing the fuel assembly into the transport device, and transporting the fuel assembly horizontally from the reactor to the fuel pool.

When installing the fuel assembly transport device in position to remove the fuel assembly, a gas sample is taken from the volume within said transporting device immediately prior to placing the fuel assembly into the transport device, and the background values of β- and γ-activity are determined before taking samples from particular fuel assemblies. Background radioactivity values are recorded for subsequent fuel assembly leakage detection analysis.

The fuel assembly is placed into the fuel assembly transport device. Placing the fuel assembly into the transport device includes gripcellg the fuel assembly and elevating it into transport position, in which the horizontal transport of the fuel assembly is possible. Usually, a rod of the fuel assembly reloading device is used as a fuel assembly transport device.

Once the fuel assembly is retreaved into transport position, a gas is supplied under the fuel assembly, thus performing bubbling, i.e. passing said gas through the heat transfer liquid. The heat transfer liquid is usually water. Preferably, the procedures of taking a gas sample from the volume above the level of heat transfer liquid inside the fuel assembly transport device and the process of fuel assembly elevation are commenced simultaneously. This allows to determine backrgound values of β- and γ-activity during the initial phase of fuel assembly elevation and to control the sample activity thereinafter. In another embodiment of the present method, the procedures of supplying the gas and taking a sample are commenced after a predetermined time interval after placing the fuel assembly into the transport device. In this case, said time interval is used to ensure that all of the gas supplied under the fuel assembly passed through the heat transfer medium, and, consequently, if non-tight fuel cells are present in the fuel assembly, that the maximum amount of gas fission products passed into the volume above the heat transfer medium and was taken as a sample. Said embodiment allows to increase precision of fuel assembly leakage detection. In the last embodiment of the present method, the procedure of taking a sample is stopped after determining the background radionuclide content within the transport device and before the predetermined time interval is over.

Sampling is performed at least at one point in the gas volume above the surface of heat transfer liquid inside the fuel assembly transport device.

Prior to measuring β- and γ-activity, the sample is prepared; said sample must conform to temperature, humidity and pressure conditions required to provide quality operation of measuring equipment and to achieve measurement error values within the margin of error of said equipment. Preparing the sample includes cooling, drying and filtering stages. Furthermore, the sample representativity control is performed. The sample is considered representative if it was taken at a determined point, it is not mixed with environmental air, and it conforms to temperature, humidity and pressure conditions. Control of the first two criteria is performed prior to reloading and control procedures by checking equipment status. Conformity control is performed using corresponding sensors when taking and analyzing the sample. If the sample does not conform to any of the conditions, a message stating that measurements were taken from a sample not conforming to working conditions of measuring equipment is displayed using the indication means. The measurement of β- and γ-activity of the gas probe is preferably performed using a beta radiometer within a time interval preset during equipment setup. The measured values are preferably displayed using available indication means and preferably saved in a text file. If the value of γ-activity exceeds the preset treshold value, it is presumed that increased β-activity of the sample may be caused by external factors, and not by fuel cell failure. Said threshold value is preferably set based on γ-radiation obtained during preliminary determination of overall background radionuclide content. The background γ-activity measured prior to fuel assembly leakage control is also taken into account. β- and γ-activity of the sample are measured, and the resulting values are recorded.

When analyzing the measurement results, it is preferable to take into account the background β-activiry value measured prior to placing the fuel assembly into the transport device, as well as the overall background β-activity measured in advance, prior to starting reloading procedures.

The fuel assembly is tentatively considered tight if the arithmetical mean of measured β- activity values in the gas sample for the current fuel assembly in the preset time interval does not exceed the predetermined threshold value; otherwise, the fuel assembly is considered non-tight. The received control result is referential, the validity thereof is determined by the validity of setting the threshold value. In one embodiment of the present invention, said threshold value is set based on overall background radionuclide content value prior to starting the reloading procedures and on test data accumulated over the operational life of the reactor. In another embodiment, the threshold value is equal to three times the background β-activity value measured prior to placing the fuel assembly into the transport device. Said methods of determining the radionuclide content threshold value during preliminary fuel assembly leakage detection are provide by the way of example and are not to be considered limiting features of the present method. The tentative result of fuel assembly leakage detection is displayed using indication means, thus allowing the nuclear power plant staff to immediately direct the non-tight fuel assembly for further detection in FFDS canisters without waiting for the reloading procedures to finish.

The horizontal transport of the fuel assembly is performed only after supplying gas under the fuel assembly in order to eliminate a chance of gas or part thereof not passing into the above-liquid volume from which the samples are being taken. Performing the horizontal transport of the fuel assembly after the supply of gas is complete allows to increase reliability of fuel assembly leakage control.

All actions in the disclosed method for fuel assembly leakage control are performed sequentially for each fuel assembly; the statistical processing of all analytical results pertaining to samples from all fuel assemblies is then carried out. In the preferred embodiment of the present method, the arithmetical mean is calculated from radioactivity values of gas samples determined during detection cycle of each fuel assembly, and then the standard deviation of said values is calculated. Hereinafter, the "detection cycle of the fuel assembly" refers to a sequential execution of actions according to the present method for one fuel assembly.

The fuel assembly is considered tight if the radioactivity value determined in a detection cycle of said fuel assembly does not exceed the threshold value, which is calculated as a sum of said arithmetical man and the tripled standard deviation value. It should be noted that said statistical processing method is provided by the way of example and is not to be considered a limiting feature of the present method. Any method can be used for statistical processing of the measurement results and determining the fuel assembly screening criterion, as long as it meets the requirements for reloading and fuel leakage detection for the particular reactor. The object of the invention is further solved by a system for fuel leakage detection of the fuel assemblies in a reactor with heat transfer liquid, adapted to be installed in a fuel assembly reloading device. The system is used for detecting fuel assemblies with non- tight fuel cells in a stopped reactor based on radioactivity of gas fission products emitted into the heat transfer liquid filling the inner chamber of the fuel assembly transport device.

Said system comprises:

- a gas supply pipeline placed on the outer surface of the outer section of the reloading device rod, comprising an injector unit under the end part of the outer section of the rod, - a gas sampling pipeline placed on the outer surface of the outer section of the reloading device rod, inserted into the outer section of said rod at at least one point,

- a gas supply unit connected to the gas supply pipeline;

- a gas sampling, preparation and detection unit connected to the gas sampling pipeline; - a control and data processing unit;

- remote control equipment.

In the system of the present invention, the means for gas sample radioactivity control and means for preliminary data processing and control are placed directly on the fuel assembly reloading device. In the preferred embodiment of the present invention, the injector unit includes injectors arranged in an annular manner to provide uniform air supply under the fuel assembly and into the space surrounding the fuel assembly and filled with heat transfer liquid. The injector nozzles can be advantageously formed as Laval nozzles, which allows to increase air spray distance at a predetermined pressure. Such arrangement allows to provide the supply of a maximum amount of gas under the back end of the fuel assembly and to expel a maximum amount of gas fission products (if non-tight fuel cells are present) from the heat transfer liquid into the volume above the surface thereof, thus increasing precision of fuel assembly leakage detection. The annular injector unit can be preferably made removable for convenience of maintenance. The gas supply unit comprises a receiver, a compressor for pumping gas into the receiver, a gas supply pressure controller and valves for controlling gas supply and for dumping condensate. Compressed air can advantageously be used as said gas.

The gas supply unit is adapted to perform an additional function of blowing down the volume above the surface of the heat transfer liquid in the rod in order to remove the sample residua of the previous fuel assembly; to that end, the unit output is connected to the sampling pipeline via a valve.

The system of the present invention is further characterized in that the gas sampling, preparation and detection unit contains means for sample conditioning. The gas sampling, preparation and detection unit also comprises the gas sample radioactivity analyzer. A beta radiometer can be advantageously used as the analyzer.

The system is further characterized in that the gas sampling, preparation and detection unit contains at least two pumps, one of which is a vacuum pump used for delivering the sample to the unit input, and the second pump is used for pumping the sample through said means for sample conditioning and said analyzer.

The means for sample conditioning includes a cooler, at least one filter, and a dryer.

Furthermore, the gas sampling, preparation and detection unit comprises at least one pressure controller, at least one pressure sensor, at least one temperature sensor and at least one humidity sensor, wherein said pressure controller is used to provide the necessary gas consumption and pressure at the beta radiometer input, and said sensors are used to determine sample representativity. The gas sampling, preparation and detection unit further comprises at least one air consumption sensor used to control the passage of sample through the beta radiometer chamber in a sufficient volume. The control and data processing unit comprises computing hardware means (a computer), signal converters, indication means, control devices and means for communication with the remote control equipment.

Advantageously, the present system provides the possibility of preliminary determining fuel leakage upon finishing a detection cycle of a single fuel assembly. The indication means of the control and data processing unit are advantageously used to output the preliminary result.

In the preferred embodiment of the system, the control and data processing unit comprises a graphical interface with data regarding fuel assembly detection cycle process, in particular, readings of pressure sensors, temperature sensors, humidity sensors and beta radiometer, displayed thereon.

The control and data processing unit is used to switch the operational modes of the present system, as well as to form control signals corresponding to said modes for the compressed air supply unit and the gas sampling, preparation and detection unit. In the preferred embodiment of the present invention, it is possible to set measurement time, threshold value of sample radioactivity, to select the operational mode of the present system, as well as to input additional information such as operator ID and the service number of the controlled fuel assembly, using control devices during setup. The control device is preferably a control panel comprising a keyboard, buttons and switches. The remote control equipment allows the personnel stationed outside the reactor hall to control the fuel assembly leakage detection. In the preferred embodiment, the remote control equipment comprises indication means, control devices, computing hardware means and data storage means; said equipment is placed in the reloading device control panel building [?] and is connected to the control and data processing unit via data channel. A conventional data channel, such as RS-422, can be advantageously used.

In the preferred embodiment, the system is adapted to be connected with an external control device. To this end, the remote control equipment comprises a standard interface that can be connected to an external device if needed, e.g. for connecting the control system of the reloading device. The connection and data transfer from the external system can be advantageously performed via Ethernet.

In the preferred embodiment, the system is adapted to be operated in at least three modes: remote-controlled automated mode, automatic mode and manual mode.

The manual mode is used to control individual devices of the compressed air supply unit and the gas sampling, preparation and detection unit in order to ensure that said units are operational. The detection cycle start-up is blocked in this mode. In manual mode, the control over devices is performed by means of control devices of the control and data processing unit, and the control over system parameters is performed via indication means of said unit.

The automated mode is the main system operation mode. In said mode, the operator can choose to control individual system devices or to initiate the fuel assembly detection cycle.

Depending on the location from which commands originate, the mode is subdivided into "remote control automated mode", wherein commands are issued via remote control equipment, and "local control automated mode", wherein commands are issued directly via the control and data processing unit. In the "remote control automated mode", the control commands are formed by the remote control equipment, and the control and data processing unit serves to transmit commands and data from the remote control equipment to the gas supply unit and the gas sampling, preparation and detection unit devices and to transmit data from said devices to the remote control equipment. In this case, operation of system devices via control devices of the control and data processing unit is blocked. The initiation of fuel assembly leakage detection cycle is executed via remote control equipment. It is advantageous to utilize such indication means and control devices as graphical interface, keyboard and mouse for the convenience of the operator working with remote control equipment. In this case, the control over the fuel assembly detection cycle is carried out by computing hardware included in the remote control equipment. The preliminary result of fuel assembly leakage detection, sensor readings and other information regarding the ongoing fuel assembly detection cycle is displayed using indication means of the remote control equipment, e.g. it is displayed on the graphical interface and duplicated on indication means of the control and data processing unit. In the "local control automated mode", the control commands are formed by the control and data processing unit, and the remote control equipment only duplicates the information on indication means. The fuel assembly detection cycle is initiated using control means of the control and data processing unit. In the "local control automated mode", the control over the fuel assembly detection cycle is performed by computing hardware means included in the control and data processing unit. The preliminary result of fuel assembly leakage detection, sensor readings and other information regarding the ongoing fuel assembly detection cycle is displayed using indication means of said unit, e.g. it is displayed on the graphical interface and transmitted to the remote control equipment for indication. In the "local control automated mode", the control commands from the remote control equipment are blocked.

The automatic mode is a special case of remote control automated mode. In said mode, the fuel assembly leakage detection cycle is initiated by the external control system, in particular, the reloading device control system. The automatic mode can be carried out if said external system is provided with a corresponding interface for connecting to the remote control equipment of the system according to the present invention. In the automatic mode, data regarding the service number of the controlled fuel assembly, operator ID and the fuel assembly leakage detection cycle initiation signal are transmitted from the reloading device control system to the remote control equipment. Said data exchange is advantageously carried out via a conventional data channel, e.g. via Ethernet.

In the preferred embodiment, the present systemcan be protected from unauthorized access to functional capabilities thereof and to data acquired during system operation. In the preferred embodiment of the invention, said protection comprises at least three clearance levels, tentatively named "operator", "engineer", "system manager". At each clearance level, it is possible to access a set of functions defined for said clearance level. It is preferable to provide protection from unauthorized access both in the control and data processing unit and in the remote control equipment. The technological outcome of using the method and system of the present invention lies in increasing reliability of fuel assembly leakage control and reducing time required to perform the fuel leakage control procedure.

In particular, blocking the horizontal transport of the fuel assembly prior to supplying gas under the fuel assembly, sample conditioning and representativity control, and taking into account background β- and γ-activity values allow to increase reliability of fuel assembly leakage control. Outputting the preliminary detection result of each fuel assembly allows to reduce time consumed by the fuel leakage detection procedure. Furthermore, the possibility of selective fuel assembly control in FFDS canisters based on the preliminary result obtained using the method and system of the present invention, allows to reduce the amount of liquid radioactive waste and to conserve material resources used in FFDS canisters. The possibility to reserve control over the fuel assembly leakage detection system, which provides the possibility to perform control using both remote control equipment and control devices located directly on the reloading device, increases system reliability in case of communication channel leakage and provides further comfort of operation for the nuclear power plant personnel. One embodiment of the present invention is further described in greater detail.

Fig. 1 illustrates a flow chart of the fuel leakage detection system.

Fig. 2 illustrates a simplified pneumatic diagram of the fuel leakage detection system.

As schematically shown in fig. 1 , the compressed gas supply unit, the gas sampling, preparation and detection unit, the control and data processing unit, as well as the air supply pipeline and the sampling pipeline are located directly on the reloading device. Said pipelines are located on the outer surface of the rod, which functions as a fuel assembly transport device. The gas supply pipeline is connected to an annular injector unit placed at the end of the outer section of the rod. In this example, the injector unit is connected to the gas supply pipeline with a conical coupler, and is screwed to the end of the outer section of the rod; for simplification, said connections are not shown in fig. 1. In other embodiments of the present invention, any method of connection can be used that provides the possibility of injector unit disengagement without dismantling the whole rod. Injectors formed as Laval nozzles provide uniform air supply under the fuel assembly with maximum spray distance of supplied air. The annular arrangement of injectors at the end of the outer section of the rod does not limit movement of the inner sections of the rod for removing the fuel assembly.

The gas sampling pipeline enters the inside of the rod at three points. Fig. 1 illustrates only one sampling point for convenience. Sampling points are placed near the surface of the heat transfer medium. Sampling at three points allows to minimize the effect of random factors on the result of determining radioactivity of the gas sample taken from the volume above the surface of the heat transer liquid.

The compressed gas supply unit, the gas sampling, preparation and detection unit, and the control and data processing unit are formed integrally, as a modular bay. The control panel comprising a keyboard, a graphical interface and control buttons for individual devices, is placed on the front face of the modular bay.

The remote control equipment is placed in the reloading device control panel building [?], thus allowing the operator to be stationed outside the reactor hall.

The control and data processing unit is connected to the remote control equipment by a data channel (RS-422 type) in order to receive commands and transmit the fuel assembly detection cycle results to record said data in corresponding data storage devices of the remote control equipment and to perform subsequent statistical processing. The control and data processing unit is connected to the compressed air supply unit and the gas sampling, preparation and detection unit by means of signal converters, input/output modules, contactors and on-off devices not shown in fig. 1 , to control said units by forming control commands in accordance with the disclosed method for fuel assembly leakage detection and to obtain detection results.

In this embodiment, the possible variations of remote control equipment and the control and data processing unit are not disclosed in detail. Any hardware or hardware/software solution can be implemented as long as it provides the functionality of said remote control equipment and said control and data processing unit.

As shown in fig. 2, the compressed air supply unit and the gas sampling, preparation and detection unit are functionally independent devices sharing a common pneumatic line from an air-control valve 5 to a pipe junction "Sample/Blowdown". The common pneumatic line allows to use the sampling pipeline connected thereto for sampling as well as for blowing down the above-water volume of the rod in order to remove residual fission products after the fuel assembly detection cycle is complete.

A compressor 1 in the compressed gas supply unit pumps the gas (in this case, air) into a receiver 2. The air is then supplied to the compressed gas supply unit output through a filter/pressure controller 3 which simultaneously dries the air and stabilizes pressure thereof at the unit output. The compressed gas supply unit is connected to the gas supply pipeline via an air-control valve 4 and a pipe junction "Sample". In order to perform bubbling, the air-control valve 4 is opened, and the compressed air is supplied under the fuel assembly via the gas supply pipeline connected to an injector unit. The bubbling duration is set using software settings, in this case said duration is about 10 seconds.

After the fuel assembly detection cycle is complete, prior to commencing the detection cycle of the next fuel assembly, a blowdown of the above-water volume of the rod is performed (if necessary) in order to remove gas fission products. In order to blow down the above-water volume, air is supplied from the gas supply unit through the air-control valve 4 switched to blowdown mode and the pipe junction "Sample/Blowdown" to gas sampling pipeline.

The control over air-control valves 4 and 5 is performed using an electrical signal from the control and data processing unit.

In order to take a gas sample from the above-water volume, the air-control valve is switched to sampling mode to connect the gas sampling, preparation and detection unit to the gas sampling pipeline. A vacuum pump 7 is therefore used to pump the gas sample from the above-water volume of the rod, first passing said sample through the water separator 6 for vacuum conditions. Then the gas sample is pumped using a pump 18 through the following sample conditioning means: cooler 8, microfilter 9, coalescing filter 10, and air dryer 1 1. The microfilter 9 and the coalescing filter 10 are used for a two- phase sample purification in order to prevent contamination of the following dryer 1 1. Then the sample is passed through a pressure controller 12 and a metering valve 13 that serve to set up the required gas consumption and pressure as it enters the beta radiometer 16 input. The sample is thus prepared, i.e. it is brought to conditions wherein the temperature, pressure and humidity thereof conform to the requirements of the measurement devices, in this case, the beta radiometer. A pressure sensor 14 and a temperature and humidity sensor 15 control the gas sample condition at the beta radiometer 16 input, and the air consumption sensor 17 is used to control the passage of sample through the beta radiometer chamber . The gas sample is released into the environment via the pipe junction "Exit". Pumps 7 and 18 are switched on for sampling using an electrical signal from the control and data processing unit (not shown in fig. 2). Signals from all sensors of the gas sampling, preparation and detection unit, as well as beta radiometer readings, enter the control and data processing unit, wherein the signals are converted and subsequently processed by the unit computer in order to determine the condition of controlled fuel assembly. The graphical interface and indicators of the control and data processing unit and of the remote control equipment (not shown in fig. 2) display information pertaining to the current condition of the system. In particular, if sensor 14 and 15 readings do not conform to set conditions, a message is displayed on the graphical interface stating that measurements are performed on a sample not conforming to measurement conditions. Furthermore, said graphical interface and indicators display the service number of the controlled fuel assembly, the operator ID, the control mode and clearance level, current sensor and beta radiometer readings, and preliminary fuel leakage detection result after the detection cycle of said fuel assembly is complete.

In the disclosed embodiment, the method and system of the present invention are used as an addition to the main procedure of fuel cell leakage control in FFDS canisters performed in a nuclear power plant. A quick screening of fuel assemblies based on the preliminary fuel leakage detection result obtained after the detection cycle of each fuel assembly is complete, allows to carry out a selective fuel assembly control in FFDS canisters without waiting for the reloading procedures to finish, which reduces the reactor idle time associated with the fuel leakage detection procedure. Furthermore, high reliability of detection results helps to prevent the fuel assemblies with non-tight fuel cells from entering the reactor.