RU2811570C1 - Method for determining parameters of volumetric distribution of nuclear fuel in stack of stopped uranium-graphite reactor - Google Patents

Abstract

FIELD: volumetric distribution of nuclear fuel.

SUBSTANCE: invention relates to a method for determining the parameters of the volumetric distribution of nuclear fuel in the stack of a stopped uranium-graphite reactor. In order to search for areas with fuel spills, the intensity of thermal and epithermal neutrons in cells in a graphite stack is remotely measured. Determination of the shape and size of the volumetric distribution zone in which fuel spills are detected is carried out by continuous measurements of epithermal neutrons along the height of the cell measurements from the starting cell in eight directions every 45 degrees, for which the cell with the maximum values of the intensity of epithermal neutrons relative to the surrounding cells is selected. By analysing the distributions of epithermal neutrons in cells, the boundaries of the zone are determined in which the intensity values of epithermal neutrons are approximately equal and do not differ by more than 30% from the average value. The graphite stack zone with a uniform distribution of the epithermal neutron field is described by an equivalent cylindrical volume, and the nuclear fuel distribution parameters are calculated.

EFFECT: ability to remotely determine the location, size and mass of the volumetric distribution of nuclear fuel in a stack of spills of fuel fragments in reactor cells, which makes it possible to assess nuclear safety.

1 cl, 7 dwg

Description

The invention relates to nuclear physics, namely to methods of radiation inspection of shut down uranium-graphite reactors, and can be used to detect spills of fragments of irradiated nuclear fuel remaining in the graphite stacks of nuclear reactors after the cessation of their operation.

The graphite masonry of a nuclear reactor consists of blocks assembled into vertical columns that penetrate vertical channels. The number of columns can reach several thousand. Spills of nuclear fuel fragments in graphite masonry are finely dispersed fractions of fuel and their mixture with graphite dust, localized on the surfaces of the blocks. Some spills of fuel fragments during operation spread throughout the volume of some zones of the graphite masonry, including tens and hundreds of graphite columns of blocks, thereby forming a volumetric distribution. The radionuclide composition of nuclear fuel fragments is characterized by the presence of a neutron source ²⁴⁴ Cm and gamma radiation sources of fission products - ¹³⁷ Cs, ¹⁵⁴ Eu, etc. To decommission reactors, it is necessary to determine the parameters and zones of their distribution of nuclear fuel fragments to assess nuclear and radiation safety.

There is a known method for detecting and determining the parameters of nuclear fuel fragments in the masonry of a stopped uranium-graphite reactor [RU 2649656 C1, IPC G21C 17/00 (2006.01), publ. 04/05/2018], selected as a prototype , which includes measuring the distributions of gamma and neutron radiation along the height of channels in columns consisting of blocks. Scanning was carried out in a certain sequence and in steps along the height of the channels. Subsequently, the distributions obtained by height are processed using mathematical algorithms, which make it possible to decompose the distribution of neutron radiation into its individual components. Calibration measurements were performed using two fast neutron sources. In this case, the registration efficiency coefficients were determined only in the neighboring cell.

Disadvantages of this method:

- the algorithm for decomposing the distribution of neutron radiation proposed in the method is intended to analyze the point localization of fuel fragments in joints and surface defects and does not allow determining the parameters of nuclear fuel fragments with their volumetric distribution in a certain area of the masonry;

- measurements of neutron radiation in height are carried out with a certain step and time delay at the measurement point (from 30 seconds or more), which, when scanning a large area of masonry, will take a lot of time and will lead to unnecessary dose loads on personnel performing work in the nuclear reactor hall.

There is a known method for detecting nuclear materials in the ground and a model for testing the method [RU 2262724, IPC G01V 5/04, publ. 10/20/2005], chosen as an analogue . According to this method, in a selected well located near the expected location of nuclear materials, the neutron field is studied depending on the immersion depth, azimuth and spectral composition of neutrons. Based on an analysis of the calculation results using programs and constants previously verified in experiments on a mock-up, all the characteristics of which are known, the size, shape and location of nuclear materials in the ground are estimated.

Disadvantages of this method:

- measurements of only thermal neutrons are performed; measurements of gamma radiation and epithermal neutrons are not provided in the cells, which make it possible to obtain additional information about the geometry of the propagation zone of fuel fragments and, accordingly, increase the accuracy of estimates;

- the thermal neutron detector uses a collimator with a hole, which must be directed strictly in one selected direction when lowering the detector along the height of the well. In this case, lowering the detector on the cable, without using additional fixing devices, can lead to rotation of the detector and, accordingly, inaccuracies;

- it is necessary to develop and manufacture a special layout for data verification, which significantly increases the research time and reduces accuracy in case of errors in reproducing identical soil and measurement conditions.

There is a known method for detecting fragments of nuclear fuel and determining their parameters in a graphite block of a nuclear reactor [RU 2 798 506, IPC G21C 17/00, G01T 1/167 17/00, publ. 06/23/2023], chosen as an analogue . Using this method, measurements of the intensity of gamma radiation, thermal and suprathermal neutron radiation from the characterized extracted graphite block are performed, including assessment of the size, shape and location of nuclear fuel fragments based on the analysis of experimental and calculated results using pre-verified programs and constants. Measurements are performed by detectors installed in the gripper and measuring assembly during manipulation of a graphite block removed from the graphite stack.

Disadvantages of this method:

- the parameters of nuclear fuel fragments are determined for only one graphite block removed from the stack;

- a significant number of detectors are involved in the measurement process, which leads to an increase in the cost of the implementation method and the risk of equipment failure.

The technical result of the proposed invention is a method for remotely determining the parameters of the volumetric distribution of nuclear fuel fragment residues in the masonry areas of a nuclear reactor, including assessment of the size, shape, location and mass of nuclear fuel.

To achieve the specified technical result, the method for determining the volumetric distribution of nuclear fuel fragment residues, as well as in the prototype, provides for search measurements of the intensity of thermal, epithermal neutrons and gamma radiation in cells to detect and determine the location of zones with fuel spills and calibration measurements.

In this case, the method differs in that scanning along the height of the cells is carried out in a wide area of spill localization and at the same time the degree of uniformity of the distribution of neutron radiation intensity is assessed. The calibration coefficient is determined by performing measurements with a superheat neutron detector depending on the distance from a point certified source of fast neutrons. Calibration measurements are performed directly in the area of the graphite stack without spillage of fuel fragments, which includes at least 10 cells.

Determination of the shape and size of the area in which fuel spills are detected is carried out by taking measurements from the starting cell in eight directions every 45 degrees. Based on the results of search measurements, the cell with the maximum value of the epithermal neutron intensity is selected as the starting cell. If measurements in a nearby cell show higher values, then it is selected as the starting one and scanning is performed according to the same pattern, with the scanning area expanding. Scanning the cells in height continues until the values of the epithermal neutron detector are higher than the background values. Determination of the final boundaries of the zone, including size and shape, is carried out by analyzing the results of measurements of the distribution of the neutron flux and identifying zones with a uniform distribution of the neutron field. To do this, determine the zone of cells in which the intensity values of suprathermal neutrons N _i are approximately equal and do not differ by more than 30% from the average value N _avg, which is determined by the formula:

, (1)

Where:

n is the number of measurements of suprathermal neutron intensity performed;

N _i - epithermal neutron count rates, measured along the height and radii of the zone.

Next, the zone, with approximately equal values of N _i in the radial and axial directions , is described by an equivalent volume of a cylindrical shape according to the formula:

(2)

Where:

h is the height of the zone , r is the equivalent radius of the zone described by the circle.

The equivalent radius r is determined by the formula:

(3)

Where:

l is the width or length of an equilateral graphite block,

π is the number "pi".

The concentration of fuel spills in the volume of each zone C _u is determined by the formula:

(4)

Where:

N _Wed - average value of intensities of suprathermal neutrons in the volume of cells of the zone with fuel spills;

α is the specific neutron yield from fuel spills;

- correction factor for conversion to the volumetric geometry of a cylindrical zone;

k - calibration coefficient.

The calibration coefficient k is determined taking into account the parameters of calibration measurements from a fast neutron source using the formula:

, (5)

Where:

τ is the age of neutrons in graphite for a certain neutron energy (reference data);

d is the distance from the detector to the source of fast neutrons in graphite;

Ngrad. - detector count rate from a fast neutron source;

S is the neutron yield from the fast neutron source.

The mass of fuel spills M in each zone is determined by the formula:

To calculate the total mass in the masonry, the obtained mass values in the cell zones are summed up.

Thus, the proposed invention makes it possible to determine the parameters of the volumetric distribution of nuclear fuel fragment residues in the masonry of a nuclear reactor, including the size of zones, shape, location and mass of nuclear fuel in the graphite masonry of a stopped uranium-graphite reactor.

In fig. Figure 1 shows an algorithm for determining the parameters of the volumetric distribution of nuclear fuel in the masonry.

In fig. Figure 2 shows a diagram of the arrangement of scanning device blocks in the central hall of a nuclear reactor.

In fig. Figure 3 shows a diagram of calibration measurements in graphite masonry.

In fig. Figure 4 shows a diagram of a detection unit with increased registration efficiency.

In fig. Figure 5 shows a zone with scanning directions relative to the starting cell and circumscribed circles.

In fig. Figure 6 shows a graphite stack zone described by an equivalent cylindrical volume.

In fig. Figure 7 shows the distributions of thermal and suprathermal neutrons from a point source of fast neutrons depending on the distance.

The method for determining the parameters of the volumetric distribution of nuclear fuel in the stack of a stopped uranium-graphite reactor is performed in accordance with the algorithm presented in Fig. 1.

At the first stage, calibration measurements are performed to determine the dependence of the detection efficiency of thermal and epithermal neutron detectors on the distance to the point source. For this, as can be seen in Fig. 2, on the upper deck of the reactor 1, a scanning device is installed (Fig. 2), consisting of a control panel 2 and a scanning unit 3, connected by a communication line 4. The scanning unit 3 is installed above the channel on the upper structures of the reactor for the duration of measurements. The channel scanning process is controlled by

remotely through control panel 2, which is a computer with special software, which makes it possible to ensure radiation safety for personnel. One of the zones 5 of the graphite stack without fuel spills is selected, which is confirmed by the absence of neutron radiation in the cells. According to the diagram presented in Fig. 3, calibration measurements are performed using a certified fast neutron source. A source of fast neutrons 6 (Cf - 252) is installed in cell I at approximately half-height of the graphite stack and, starting from the neighboring cell II through VI , the intensity of epithermal neutrons is then measured in positions 8 while the values of the detector 7 are higher than the background values of the detector. Thermal neutron measurements are performed in the same order.

Next, the obtained distributions of suprathermal and thermal neutrons are analyzed and the calibration coefficients for the detection efficiency of the neutron detector are determined using formula (5) and the maximum distance at which this type of detector can detect a neutron source in continuous scanning mode. Next, this distance between cells is used as a step for search scanning.

At the second stage, a search neutron scan is performed in the area of cells of the graphite stack of a uranium-graphite reactor by measuring the fluxes of thermal and epithermal neutrons with a previously determined step between cells. To ensure continuous scanning, a neutron detector 7 with increased neutron detection efficiency is used. To increase the efficiency of neutron registration, the detector (Fig. 4) is made of a counter assembly (at least four) 9 and an electronics assembly 10 based on a signal conversion unit, which makes it possible to increase the neutron registration efficiency of the detector is proportional to the number of counters 9 and to move away from the step mode when the detector stopped after a certain time in order to collect sufficient neutron registration statistics due to the insufficient registration efficiency of a detector based on a single counter.

A removable cadmium cover 11 is installed on the detector to register epithermal neutrons, which is removed to register thermal neutrons. Also, to increase efficiency, a neutron moderator made of polyethylene 12 is installed in the detector. Using an automatic scanning unit, to which it is connected by a cable through connector 14, the detector is moved along the height of the channel in a continuous mode. Readings from the detector are taken in an automatic non-stop mode of movement every 10 mm in height.

Based on the results of search measurements, a zone containing fuel spills is determined, in which a cell with the maximum intensity of epithermal neutron radiation is determined, relative to neighboring cells, which is conventionally considered the starting one 13. Next, according to the diagram in Fig. 6, cells 14 are scanned from it in eight directions every 45 degrees. According to the criteria, according to the algorithm, the uniformity of the distribution of neutron radiation intensity is determined. In this case, the differences in the values of the epithermal neutron detector in the starting and neighboring cells should be approximately equal to the average value Nav. , which is determined by formula (1), and not exceed it by more than ±30%. The criterion for classification as a volumetric distribution also includes the absence of pronounced peaks in the gamma radiation distribution, indicating the point nature of fuel localization. In the case of identifying a volume-uniform distribution of the field of epithermal neutrons, the zone is described by a cylindrical figure 15 with an equivalent radius r , then, taking into account the height of the detector readings, the height of the zone h is determined, which makes it possible to calculate the fuel mass M using formula (6) taking into account the concentration of spills fuel in the volume of the Cu zone, which is determined by formula (4). If the volumetric nature of the distribution in the zone is not identified, then the calculation is performed using another algorithm, for example, considered in the prototype [RU 2649656 C1, IPC G21C 17/00 (2006.01), publ. 04/05/2018]. The value of the calibration coefficient is determined by formula (6) taking into account the dependence of the neutron distribution on the source of fast neutrons obtained according to the scheme (Fig. 3).

An example of the invention is given below.

The object chosen was the graphite stack of one of the shutdown channel uranium-graphite reactors. For scanning, an automated scanning device was used, equipped with a detector with increased registration efficiency based on a counter unit. For scanning, we used a radiation-resistant detection unit containing an assembly of four neutron counters, a polyethylene moderator, and a removable 1 mm thick cadmium cover. To register gamma radiation, a radiation-resistant gamma radiation detection unit with a Si detector was used.

At the finally stopped channel uranium-graphite reactor, cells were scanned in a certain order according to the algorithm (Fig. 1). In fig. Figure 7 shows the distributions of suprathermal 16 and thermal 17 neutrons from a point source of fast neutrons obtained during calibration measurements depending on the distance. It was determined that at background values of the detector ~10 pulses per second

Using a thermal neutron detector, it is enough to search with a step of 2400 mm between cells. Next, from the cell in the zone with spills, chosen as the starting point, measurements were taken of the distribution of epithermal neutrons in eight directions at 45 degrees in the zone and the height of the cells. When higher counting rates were received from the detector in neighboring cells, the choice of the starting cell was adjusted. After detecting uniformly distributed fields of suprathermal neutrons in the zones, which, taking into account the criteria presented in the algorithm, indicated the volumetric nature of the fuel distribution, the mass was determined using formulas (1-7). Calculations of radionuclide accumulation taking into account mass values²⁴⁴Cm was performed using the verified WIMS-D4+CACH-2 program, which allows one to calculate the dependences of the accumulation of Cm and other radionuclides and the specific neutron yieldα on the time of irradiation and exposure. Dependency data showing accumulation relationships²⁴⁴Cm relative to nuclear fissile radionuclides Pu, Am etc., allowed based on known mass²⁴⁴Cm in the characterized block zone also determine their mass.

Thus, the implementation of the proposed invention makes it possible to accurately determine the location, size and mass of the volumetric distribution of nuclear fuel in a stack of spills of fuel fragments in reactor cells, which makes it possible to assess nuclear safety.

Nav = 1 n Σ i = 1 n x i = 1 n ( x 1 + … + x n ).

Claims (13)

A method for determining the parameters of the volumetric distribution of nuclear fuel in the masonry of a shutdown uranium-graphite reactor, including searching and assessing the size, shape of distribution zones and mass of nuclear fuel fragments based on an analysis of experimental and calculated results using pre-verified programs and constants, characterized in that the determination of the shape and the dimensions of the volumetric distribution zone in which fuel spills are detected are carried out by continuous height measurements of cells from the starting cell in eight directions every 45 degrees, for which the cell with the maximum values of the intensity of epithermal neutrons relative to the surrounding cells is selected, and by analyzing the distribution epithermal neutrons in cells determine the boundaries of the zone in which the intensity values of epithermal neutrons are approximately equal and do not differ by more than 30% from the average value, and describe the zone of cells with an equivalent volume V _{equivalent of} a cylindrical shape according to the formula where h is the height of the zone, r is the equivalent radius of the zone described by the circle, according to the formula where l is the width or length of a graphite block equilateral in cross-section, π is the number “pi”, and then the concentration of fuel spills C _and in each zone is determined by the formula where N _cp. - average value of suprathermal neutron intensities in the volume of cells in the zone with fuel spills; α is the specific neutron yield from fuel spills; k - calibration coefficient; β - correction factor for conversion to the volumetric geometry of a cylindrical zone, the mass of fuel spills in each zone is determined by the formula then all adjacent zones are combined into one zone, the masses are summed up, and the overall dimensions and shape of the nuclear fuel distribution zone in the graphite stack of the stopped uranium-graphite reactor are determined.