86 results about "Neutron count" patented technology

The present invention provides for method and an apparatus for determining the porosity of a geological formation surrounding a cased well. The method further comprises of generating neutron pulses that release neutrons into an area adjacent the well, from a neutron source. Neutrons are sensed and a plurality of neutron detector count rates is acquired using at least two neutron source to neutron detector spacings. A timing measurement is acquired at one of the spacings to measure a first depth of investigation. A ratio of the neutron detector count rates is acquired to measure a second depth of investigation. An apparent porosity is calculated using the timing measurements and the ratios of neutron count rates. The effect of a well casing on the calculated apparent porosity is determined in response to at least one of the ratio of neutron detector count rates and the timing measurement. A cement annulus is computed based on the ratios of neutron count rates and the timing measurement. A formation porosity is calculated by performing a correction to the apparent porosity for the casing and the cement annulus.

The present invention is directed toward the determination of soil water content and the water equivalent depth of snow deposited on top of the ground. The apparatus for measuring soil water content consists of one or more cosmic-ray neutron detectors located above the soil surface. The detectors record neutrons in the thermal, epithermal or fast energy bands. Snow water equivalence is determined by using a thermal neutron detector and an epithermal or fast neutron detector. Neutron count rates for all three energy bands decrease monotonically with increasing soil water content. Epithermal and fast neutron count rates decrease monotonically with increased snow water equivalent depth, but thermal neutron count rates increase at first and then decrease with increasing snow cover.

The invention discloses a controllable-neutron three-detector element well logging device and method, which specifically relate to the field of oil and natural gas exploration. The device uses a D-T controllable neutron source, a near He-3 neutron detector, a LaBr3 crystal gamma detector, and a super far He-3 thermal neutron detector. The two He-3 neutron detectors collect the time spectrum of neutrons and the total count of neutrons. The LaBr3 gamma detector collects inelastic and capture gamma time spectrums, and measurement information is uploaded to ground equipment through a telemetry pipe nipple of the instrument. The method includes the following steps: first, processing the inelastic gamma energy spectrum, calculating the inelastic gamma yield of elements, and calculating the dry weights of Si, C and other elements according to the inelastic yield of elements, the near-to-super far neutron counting ratio and other information; and then, processing the capture gamma energy spectrum, calculating the capture yield of elements, and calculating the contents of elements which can be measured by the capture gamma energy spectrum, such as Ca, S and Fe, according to the contents of elements determined by inelastic gamma and the capture yield.

The invention discloses a method for replacing a primary neutron source with irradiated fuel assemblies. A primary neutron source assembly is not loaded or used by a reactor core; one or more irradiated fuel assemblies obtained from other running reactors are loaded in a first cycle; the neutron counting rate of a detector meets the requirements by using neutrons released by the irradiated fuel assemblies; and meanwhile, fissile nuclides preserved by the irradiated fuel assemblies meet the physical design requirements of the first cycle. The method has the beneficial effects that the fuel assemblies generate the nuclides capable of reacting with (alpha, n) to generate the neutrons through self-fission reaction after two cycles, meanwhile, a lot of fissile nuclides still exist in the fuel, and the primary neutron source is replaced with the irradiated fuel assemblies, so that the problems of Cf-252 source production, equipment manufacturing or radiation protection caused by a primary neutron source assembly are solved under the condition of meeting the reactor core loading and critical safety supervision starting requirements.

The invention relates to a method for monitoring sub-critical reactor reactivity. The method includes the steps: placing neutron detectors at two different positions of a reactor; effective multiplication constant gradually changing the reactor reactivity, and recording a specific value C and an effective multiplication constant keff of neutron counting rate at the same moment; fitting calibratingcurve graphs of the specific values C and the effective multiplication constants keff of the neutron counting rate of various moments; recording neutron counting rates of the two neutron detectors inreal time to obtain specific values of the two neutron counting rates, and substituting the specific values into the calibrating curve graphs to obtain effective multiplication constants of the effective multiplication constant when reactor reactivity is practically monitored. The neutron detectors are used for acquiring neutron counting rates, and the specific values C of the neutron counting rates recorded by two detectors represent characteristic parameters of neutron flux density space distribution shapes. Compared with a traditional method, the method is more visual, simple and convenient, on-line real-time monitoring of the reactivity change can be achieved, effects of external source intensity are omitted, and space effect correction of the traditional method under a deep sub-critical condition is avoided.

The invention relates to a grid type control drum device of a small fast neutron reactor. The grid type control drum device comprises an inner reflecting layer, a grid type control drum and an outer reflecting layer which sequentially sleeve the periphery of a reactor core from inside to outside, and a driving motor for driving the grid type control drum; the inner reflecting layer is in a cylindrical grid shape and is fixed in position; the outer reflecting layer is in a complete cylinder shape and is fixed in position; and the grid type control drum is in a cylindrical grid shape and can bedriven by the driving motor to rotate by a certain angle so as to adjust the overlapping area between the absorption grids and the reflection grid of the inner reflecting layer and change the number of leaked neutrons reflected back to the reactor core, so reactivity control is achieved. The device disclosed by the invention is small in occupied axial space, and meets the miniaturization design requirement; the absorption grids have no axial displacement, so that uneven axial power distribution of the reactor core cannot be caused; and the absorption grids are in central symmetry about the central axis of the reactor core, have good adaptability to material expansion caused by radial temperature gradient, and are not easy to block due to deformation.

The invention relates to a fast-neutron multiple measuring-analyzing method based on a liquid scintillation detector. The method comprises the following steps of building a mathematic model of fast-neutron multiple measurement based on the liquid scintillation detector, wherein the mathematic model is an equation set which is related to various multiple coincidence counting rates and unknown sample parameters; employing a plurality of liquid scintillation detectors to detect the sample source neutron fission process so as to obtain a neutron counting rate, and further obtaining various multiple coincidence counting rates; substituting the measured various multiple coincidence counting rates into the equation set, and solving the equation set so as to obtain the unknown sample parameters (spontaneous fission rate and induced fission rate). By employment of the fast neutron multiple measuring-analyzing method based on the liquid scintillation detector provided by the invention, a stable and reliable analysis result of a measured sample can be given out.

The invention relates to a fission neutron well logging correction method, which comprises the following steps that: 1, a correction counting rate of a neutron monitor of a tested wellhole is obtained; 2, a water correction factor is obtained; 3, the neutron counting rates near a well wall and along a well axis after the water correction in a tested well are obtained; and 4, the neutron counting rates near the well wall and along the well axis after the epithermal neutron service life correction in the tested well are obtained. The method provided by the invention can be used for directly measuring the uranium content. Compared with drill hole rock core sampling chemical analysis results, the method has the advantages that the deviation does not exceed 10 percent. The uranium ore grade is more objectively reflected; particularly in sections with damaged uranium-radium balance, conventional gamma well logging is combined, and a uranium-radium balance factor of a drill hole ore block can be determined in site; and the rock core sampling quantity is reduced, and the uranium resource investigation cost is reduced. The method can be used for accurately measuring the uranium leaching rate in mining using the leaching technology, avoids the environment pollution and the cost waste due to unquestioning excessive liquid injection in the process of the mining using the leaching technology, and has important practical significance on improving the efficiency of the mining using the leaching technology, reducing the cost and protecting the environment.

The invention belongs to the technical field of neutron detection, and provides a concrete emptying detection method, device, and terminal equipment. The method comprises following steps: reading thethermal neutron counting rate N1 of a first thermal neutron detector and the thermal neutron counting rate N2 of a second thermal neutron detector, calculating the counting rate ratio of N1 to N2, wherein the distance between the first thermal neutron detector and a neutron source is smaller than the distance between the second thermal neutron detector and the neutron source; obtaining the thickness of a steel plate that covers the surface of concrete, based on the thickness of the steel plate, finding the corresponding relationship between the counting rate ratio and concrete emptying depth;and based on the corresponding relationship, determining the concrete emptying depth corresponding to the counting rate ratio. The method has the advantages that a large amount of tedious operation isreduced, and the efficiency on measuring the concrete emptying depth is largely improved.

The invention provides a measuring instrument for concrete void detection, the measuring instrument is applicable to the technical field of neutron detection, and the measuring instrument comprises aneutron source, a processing unit, a first thermal neutron detector and a second thermal neutron detector; the neutron source is used for transmitting fast neutrons to concrete covered with a steel sheet; the first thermal neutron detector and the second thermal neutron detector are used for obtaining heat neutron counting rate N1 and heat neutron counting rate N2 at two locations, with differentdistances from the neutron source, on the steel sheet covering the surface of the concrete; the processing unit is configured to calculate the counting rate ratio of heat neutron counting rate N1 to heat neutron counting rate N2; the processing unit is further configured to obtain the thickness of the steel sheet covering the surface of the concrete, according to the thickness of the steel plate,a corresponding relationship between the counting rate ratio and a concrete void depth is found, and the concrete void depth corresponding to the counting rate ratio is determined according to the corresponding relationship. The measuring instrument saves a lot of cumbersome operation and greatly improves the efficiency of actual concrete void depth measurement.

The invention discloses a method for cancelling a secondary neutron source of a reactor. A reactor core is not loaded with and does not use a secondary neutron source assembly, and neutrons released by a postdose fuel assembly loaded in the reactor core are utilized to make neutron counting rate meet requirements. The method for cancelling the secondary neutron source of the reactor has the advantages that the neutrons released by postdose fuel are utilized to meet critical safety supervision requirements during follow-up circulation reactor loading and physical start-up period; the secondary neutron source assembly does not need to be introduced again to the reactor core, and the reactor start-up process can be monitored by follow-up circulation.

The invention provides a borehole fluid time spectrum evaluation method using a PNT (pulsed-neutron tool), which includes specific steps: firstly, using a high-energy neutron generator to transmit high-energy neutrons into a stratum; secondly, using two detectors to simultaneously acquire number of low-energy neutrons at 60 intervals of from 36 microseconds to 2160 microseconds when the high-energy neutrons are transmitted from the high-energy neutron generator; thirdly, generating a neutron time spectrum according to the number of the neutrons at each time interval, converting the neutron time spectrum into a data base to be recorded, and generating imaging information according to data of the data base; fourthly eliminating impact of borehole size, a cement mantle and in-wall fluid through images according to corresponding relation of well influencing factors to thermal neutral decay so as to obtain objective stratum parameters; and fifthly outputting the stratum parameters to a display device to serve for high-quality reservoir evaluation. The data are effectively acquired, so that risks are lowered, total investment is lowered, projections of logging through casing and open hole logging are the same in type, and time shifting analysis can be performed.

The invention discloses a dual non-uniform space self-shielding effect correction method, a device, equipment and a medium. The method comprises the steps: calculating the number of neutrons escapingfrom the surface of fuel and the number of neutrons entering adjacent fuel lattice cells for collision in a moderator after escaping from the surface of the fuel through a Monte Carlo method, and obtaining the number of neutrons escaping from the surface of the fuel; obtaining a Dankov factor, and then correcting the neutron escape probability through the Dankov factor so as to correct the dual non-uniformity of neutron resonance calculation; according to the fact that the escape probability of neutrons in the randomly distributed medium area before and after homogenization is not changed, enabling the randomly distributed medium area to be equivalent to a homogenization medium; calculating the equivalent cross section of the homogenizing medium based on the macroscopic cross sections of the dispersed particles and the sub-regions of the matrix and the volume shares and the space self-shielding factors corresponding to the sub-regions, so as to correct the dual non-uniformity of neutron transport calculation and realize the correction of the dual non-uniformity space self-shielding effect of the randomly distributed medium fuel element. The neutron calculation accuracy is improved.

The invention discloses a measurement device and method of copper-nickel mineral product grades in a well. The device comprises an underground device and a ground data processing device. The underground device is connected with the ground data processing device through a cable. The underground device comprises a housing body which is in the shape of a titanium steel cylinder. A D-T controllable neutron source, a tungsten shielding body, a CLYC double-particle detector and a remote metering short section are down-up successively arranged in the housing body. The method mainly comprises following steps of firstly, according to a measured capturing gamma spectrum, selecting gamma counting information in a high energy section; calculating gamma yield of different metal elements; calculating the ratio value of the gamma yield and thermal neutron counting; eliminating effects on the gamma yield imposed by unstable fluxes of the thermal neutrons; and finally, calculating metal mineral product grades according to a linear relation between the ratio value of the gamma yield-thermal neutrons and the metal mineral product grades. According to the invention, metal mineral product grades with the continuous depths can be provided in the well; quantification of metal mineral product exploration is facilitated; and efficiency of the metal mineral product exploration is improved.

The invention discloses a hazardous article detection device and method based on neutrons and X-rays, which relate to the technical field of public safety. A control system of the device is used for controlling a conveying belt to convey a to-be-detected article into an X-ray machine to perform X-ray detection on the to-be-detected article, controlling the conveying belt to convey the to-be-detected article into a detection area of a neutron source, and controlling the neutron source to rotate through a neutron source rotating system, so that the central axis of the neutron source penetrates through a hazardous article, and controlling a neutron source and a gamma ray detector to carry out neutron detection on the hazardous article to obtain a neutron detection result; and an element analysis system is used for determining the type of the hazardous article according to the X-ray detection result and the neutron detection result. The neutron source is rotated to the position where the hazardous article, the neutron source and the neutron source rotating system are on the same line through the neutron source rotating system, so that what can be ensured is that the target piece is closest to the suspected hazardous article, the number of 14MeV neutrons at the suspected hazardous article is maximum, and the detection efficiency is improved.

A fission chamber count rate measurement device and to the associated fission chamber calibration device; the count rate measurement device comprises: (1) a measurement cell, which contains the fission chamber (CH); (2) a neutron generator, which emits neutrons in the form of periodic pulses towards the fission chamber; (3) a neutron counter (K), which detects and counts the neutrons emitted by the neutron generator; and (4) a computing circuit, which delivers, over a predetermined time interval, a fission chamber count rate normalized with reference to the number of neutrons counted by the neutron counter (K).

The embodiment of the invention provides a method, device and apparatus for measuring plutonium mass of a plutonium substance, and a medium. The method comprises the following steps of: 1, obtaining the characteristic information of each plutonium standard sample in a plurality of plutonium standard samples, and the second plutonium isotope abundance and second 241Am isotope abundance of a to-be-measured plutonium substance; each plutonium standard sample and the to-be-measured plutonium substance are measured; based on the first plutonium isotope abundance and the first 241Am isotope abundance in each piece of characteristic information, the first total neutron counting rate and the first coincidence neutron counting rate of each plutonium standard sample, and a first preset parameter, the third coincidence neutron counting rate of each plutonium standard sample is obtained; a scale curve is obtained based on the third coincidence neutron counting rate and the first equivalent 240Pueff mass of each plutonium standard sample; and according to a second total neutron counting rate, the second coincidence neutron counting rate, the scale curve, the second plutonium isotope abundance,the second 241Am isotope abundance and the first preset parameter, the plutonium mass of the to-be-measured plutonium substance is determined, and measurement precision can be improved.

The invention discloses a neutron counting method used in a strong pulse X-ray environment and a counter. The neutron counting method comprises a moderator, a proportional counting tube, an isolationresistor, an isolation capacitor, an amplification circuit, a cable and a pulse power supply circuit. The pulse power supply circuit is used for generating pulse voltage based on a pulse signal generated by the pulse generator and sending the pulse voltage to the proportional counting tube through a cable and an isolation resistor; the proportional counting tube is used for generating an originalcounting pulse signal based on the pulse voltage and sending the original counting pulse signal to the amplification circuit through the isolation capacitor, and the original counting pulse signal carries a bias signal; the amplification circuit is used for isolating bias signals carried by original counting pulse signals, effective counting pulse signals are obtained and sent to an oscilloscope so as to avoid the time interval from the time when the proportional counting tube is irradiated by pulse X-rays to generate electrons to the time before the electrons are absorbed by gas, so that theelectrons generated by X-ray irradiation ionization are not increased due to electric field acceleration, and the neutron counting accuracy is improved.

The invention belongs to the radiation measurement technical field and relates to a neutron detection method for reducing gamma ray interference by using a GM counter tube.According to the neutron detection method of the invention, when a 6LiI scintillator detector applying a 6LiI scintillator is adopted to detect neutron rays in a mixed radiation field, a first voltage amplitude discrimination threshold is set in the 6LiI scintillator detector; and the signals of low-energy gamma rays detected by the 6LiI scintillator detector are filtered out. In order to eliminate the influence of high-energy gamma rays on a measurement effect, the neutron detection method of the invention includes the following steps that: (S1) a GM counter tube is set near a 6LiI scintillator; (S2) a second voltage amplitude discrimination threshold is set for the GM counter tube; (S3) first signals measured by a 6LiI scintillator detector are recorded, and second signals measured by the GM counter tube are recorded; and (S4) the second signals are multiplied by a correction coefficient, the second signals multiplied by the correction coefficient are subtracted from the first signals, so that a net neutron count rate is obtained.

The invention belongs to the field of reactor engineering design and operation, and relates to an approximation critical method for a pressurized water reactor first furnace reactor core without an additional primary neutron source. The method comprises the following steps of: 1, installing a neutron detector monitoring device; 2, confirming the channel neutron counting response performance of a source range detector monitoring device and a high-sensitivity neutron detector monitoring device; 3, carrying out critical safety supervision parameter calculation and alarm function setting on a secondary instrument in an out-of-reactor high-sensitivity neutron detector monitoring device; 4, confirming the effectiveness of the protection function of the source range detector monitoring device; 5,adding a critical safety supervision key parameter monitoring and alarming function to a digital distribution control system of a power station; 6, heating and boosting a reactor; 7, adding a small-flow dilution control conversion point and dilution interval control; 8, isolating and controlling equipment influencing the subcritical degree of the reactor; and 9, guiding the reactor core without the additional primary neutron source to transit from a deep subcritical state to a critical state.

The invention belongs to the radiation measurement technical field and relates to a neutron detection method and device for reducing gamma ray interference by using a silicon PIN detector. According to the neutron detection method and device of the invention, a 6LiI scintillator detector is adopted to detect neutron rays in a mixed radiation field; a first voltage amplitude discrimination threshold is set in a first comparison circuit connected with a 6LiI scintillator; and the signals of low-energy gamma rays measured by the 6LiI scintillator are filtered out. In order to eliminate the influence of high-energy gamma rays on a measurement effect, the neutron detection method of the present invention includes the following steps that: (S1) a silicon PIN detector is arranged near a 6LiI scintillator; (S2) a second voltage amplitude discrimination threshold is set in a comparison circuit connected with the silicon PIN detector; (S3) first signals measured by the 6LiI scintillator are recorded, and second signals measured by the silicon PIN detector are recorded; and (S4) the second signals are multiplied by a correction factor, and the second signals multiplied by the correction factor are subtracted from the first signals, so that a net neutron count rate is obtained.