356 results about "Neutron flux" patented technology

In a method of producing isotopes in a light water power reactor, one or more targets within the reactor may be irradiated under a neutron flux to produce one or more isotopes. The targets may be assembled into one or more fuel bundles that are to be loaded in a core of the reactor at a given outage. Power operations in the reactor irradiate the fuel bundles so as to generate desired isotopes, such as one or more radioisotopes at a desired specific activity or stable isotopes at a desired concentration.

The invention provides a beam shaper for neutron-capture therapy in order to improve flux and quality of a neutron source. The beam shaper comprises a target, a slowing body adjacent to the target, a reflector wrapping the slowing body, a thermal neutron absorber adjacent to the slow body, a radiation shield arranged in the beam shaper and a beam outlet. The target generates nuclear reaction with a proton beam incident from a beam inlet so as to generate neutrons, the neutrons form a neutron beam which defines a main axis, the slowing body slows down the neutrons generated from the target to an epithermal neutron energy region, the reflector guides the neutrons deviating from the main axis to the main axis so as to improve intensity of the epithermal neutron beam, a gap passage is arranged between the slowing body and the reflector so as to improve epithermal neutron flux, the thermal neutron absorber is used for absorbing the thermal neutron so as to avoid causing overmuch dosed with shallow normal tissue during therapy, and the radiation shield is used for shielding leaked neutrons and photon so as to reduce normal tissue dose in a non-radiation region.

A method for acquiring fine distribution of the reactor core three dimensional neutron flux density of a reactor comprises the following steps: 1, carrying out geometric modeling to a reactor core, dividing computational domain, dispersing angle space, generating characteristic line, appointing materials for all computing regions, obtaining macroscopic section parameters and setting initial values for neutron flux density flux of a computed region, terminal condition of reactor and characteristic value; 2, calculating coefficient matrix required by the matrix method of each sub region and a part which is positioned at the right end and cannot be changed with the iterative computation; 3, seeking the density of thicknet neutron flux and correcting the density of one-dimensional and two dimensional thinnet neutron flux; 4, seeking the density of one-dimensional neutron flux of each lattice cell; 5, iterative solution of the density of the two dimensional thinnet neutron flux of each layer; and 6, updating three dimensional thicknet parameters, judging whether a characteristic value and the density of a three dimensional neutron flux are in convergence or not, if in convergence, turning to the step 3 for continuous iteration under convergence is achieved, namely, fine distribution of three dimensional neutron flux density can be obtained; and the fine distribution of the reactor core three dimensional neutron flux density of a reactor can be obtained quickly.

A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360° and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

Example embodiments are directed to methods of producing desired isotopes in commercial nuclear reactors and associated apparatuses using instrumentation tubes conventionally found in nuclear reactor vessels to expose irradiation targets to neutron flux found in the operating nuclear reactor. Example embodiments include assemblies for retention and producing radioisotopes in nuclear reactors and instrumentation tubes thereof. Example embodiments include one or more retention assemblies that contain one or more irradiation targets and are useable with example delivery systems that permit delivery of irradiation targets. Example embodiments may be sized, shaped, fabricated, and otherwise configured to successfully move through example delivery systems and conventional instrumentation tubes while containing irradiation targets and desired isotopes produced therefrom.

The invention relates to the nuclear reactor core power monitoring field and discloses a method for online monitoring the neutron flux distribution of the reactor core. The method bases on M1 internal neutron detectors and M2 external neutron detectors which are arranged on the reactor. According to a reference reactor core model of a renewal reactor that MPhi is equal to (1 / k) FPhi, higher order harmonics are solved. Then, reading data of the internal neutron detectors and the external neutron detectors are combined. The neutron flux distribution in the core of a real reactor is reconstructed online.

A therapy apparatus for producing thermal neutrons at a tumor site in a patient has a plurality of fast neutron sources surrounding a moderator, a fast neutron reflecting media around the fast neutron sources, a gamma-ray and neutron shielding media surrounding the fast neutron reflecting media, and a patient chamber positioned inside the moderator. The fast neutron sources are positioned around the moderator to maximize and direct the neutron flux to said tumor site.

Disclosed is a method for calculating core neutron flux distribution of a small experimental reactor. The method includes steps of 1), determining geometrical and material parameters according to a core structure of the small experimental reactor, and establishing a neutron-transport equation describing movement rules of neutrons in each discrete direction inside a reactor core; 2), preparing corresponding quadrature sets for boundaries unparallel in normal surface vectors and each coordinate axis direction; 3), subjecting the neutron-transport equation to numerical discretion to acquire simultaneous linear algebraic equations of angular neutron flux density by a segment method of approximating unstructured geometry with unstructured grids to establish arbitrary triangular grids; 4), solving the simultaneous linear algebraic equations to acquire the discrete distribution of the angular neutron flux density in the reactor core, and acquiring the discrete distribution of neutron-flux density in the reactor core by the aid of relation of the angular neutron flux density and the neutron-flux density. By the method, the neutron-flux distribution of the small experimental reactor, especially that of an isotope production reactor, a pebble bed reactor and a high-flux reactor in the medical field can be accurately acquired.

The invention relates to the fact that a common industrial neutron interrogation screening requirement is that a high throughput rate be accommodated by the screening system. The accumulation of elemental abundance ratio spectral data to minimize statistical uncertainty is a function of the neutron flux passing through the subject. If the subject passes through a neutron beam, with a strictly limited time window for exposure, the flux must be sufficient to accumulate the required statistics. The level of neutron flux necessary may exceed the cost effective limits of the selected neutron source means. Exposure time window dilation is disclosed through a class of system configurations which become practical for reduction to practice by utilization of linear neutron source topology neutron generators. This disclosure is concerned with example embodiments which utilize the length, width, thickness and segmentation of the source emission zone within an appropriate neutron source.

An incore monitoring method of a nuclear reactor, includes, measuring neutron flux levels at pitch levels corresponding to local power range monitor sensors arranged along an axial direction inside a detector assembly installed in a nuclear reactor; performing power calculation, including calculation of thermal characteristics, of fuel assembly group consisting of fuel assemblies adjacent to the corresponding detector assembly, based on indicated values of the local power range monitor sensors of the corresponding detector assembly at a first time, calculating thermal characteristics at a second, subsequent time in which the power calculation is not calculated, based on values indicated by the local power range monitor sensors and calculated thermal characteristics at the first time and values indicated by the corresponding local power range monitor sensors at the second time, and monitoring the calculated thermal characteristics.

The invention discloses a method for obtaining three-dimensional neutron flux density distribution in a reactor core transient process of a fast neutron reactor. A polygonal prism grid is adopted for the anisotropic neutron flux density of the reactor core of the fast neutron reactor to carry out fully-three-dimensional transportation space dispersion, a semi-spherical surface of a sixty-degree region is taken as a unit to carry out alternate scanning, and iterative format degradation in an angle parallel process is weakened; and the characteristics of the weak local effect and the global space coupling of the neutron flux density of the reactor core of the fast neutron reactor are considered, an estimated correction quasi-static strategy is adopted for neutron flux density change in a transient process to carry out time dispersion, ingredients of different change rates along with time in the neutron flux density can be decomposed, separated solution is carried out on different time scales, meanwhile, nonlinear iteration among above ingredients is avoided, and calculation efficiency is improved. The method for obtaining the three-dimensional neutron flux density distribution in the reactor core transient process of the fast neutron reactor has the advantages of being high in calculation accuracy and reasonable in calculated amount.

A neutron logging tool includes a neutron source and at least one position sensitive thermal or epithermal neutron detector. The logging tool further includes an electronic controller configured to estimate the axial location of detected neutrons. Measurement of the axial neutron flux distribution enables other formation and borehole parameters such as formation porosity and sensor standoff to be computed. In logging while drilling embodiments, a borehole caliper may also be computed form the axial neutron flux distribution.

The invention provides a method for precisely calculating power distribution of a fuel assembly rod. The method comprises the following steps of 1, performing resonance calculation and transportation calculation to obtain the multigroup microscopic capture cross section and microscopic fission cross section of each nuclide in each burn-up state and the netron-flux density of each calculation region; and 2, obtaining corresponding heat release energy through precise processing according to information provided by a burn-up library, and calculating the power distribution of a fuel assembly according to the obtained microscopic absorption cross section, the microscopic fission cross section, the netron-flux density and the heat release energy. The method provided by the invention has the advantage that the energy release in the fuel assembly, including energy release of fission reaction occurrence and energy release of radiation capture reaction occurrence is strictly considered, so that the precise single-rod power distribution is obtained.

An ex-core nuclear instrumentation system in which the width of measurable neutron detector current can be accurately widened is obtained. In order to output the condition of neutron flux in operation by performing arithmetic processing of a current value measured by a neutron detector by using a detector signal processing circuit, the detector signal processing circuit includes a current / voltage conversion unit which converts the current value converted by the neutron detector into a voltage value corresponding to the current value; and a variable gain amplification unit which has an operational amplifier having a resistance circuit for corresponding to current levels, the resistance circuit being capable of selecting a gain, and a D / A converter that adjusts the gain, and amplifies the voltage value converted by the current / voltage conversion unit.

The invention relates to a method and installation for monitoring the core of a nuclear reactor, and the nuclear reactor comprises a movable core-internal instrumentation system (4) whole instrumentation tube (3) enters into a nuclear reactor core (1) and a fuel instrument assembly (2') at the reactor core height so as to form a measurement channel. Each movable probe fixed at one end of each displacement cable allows measuring the flux of each neutron along the whole height of the reactor core in each measurement channel. The reactor core is monitored and measured by comparing the critical parameter of at least one reactor core with the predetermined threshold value. If the critical parameter reaches the threshold value or the value determined based on the physical value detected from the reactor core (1) departed from the value calculated based on a module, a movable probe is guide into at least one measurement channel (3) of the reactor core (1); and the movable probe moves along the height of the reactor core (1) so as to complete the flux measurement of one group of neutrons, thereby calculating the critical parameter according to the neutron flux measurement completed by the movable probe.

A reactor start-up monitoring system, comprising:a determination apparatus for determining that moderator temperature reactivity coefficient is positive based on neutron flux measured by a neutron detector and a reactor water temperature measured by a temperature detection apparatus;an output information creating apparatus for creating first output information indicating positive moderator temperature reactivity coefficient when determination information inputted from the determination apparatus indicates the positive moderator temperature reactivity coefficient; andat least one of a display apparatus and an audio output apparatus for inputting the first output information.

The invention discloses an improved method for calculating three-dimensional neutron flux density fine distribution of a reactor core. The method comprises the following steps that 1, modeling is conducted on the reactor core, needed parameters are calculated, and variables are initialized; 2, a radial leakage item is unfolded in the axial direction through a Legendre polynomial expansion, a one-dimensional discrete ordinate difference equation is calculated according to information obtained in the step 1, and an axial leakage item is acquired; 3, two-dimensional spatial distribution of the axial leakage item is obtained through calculation, two-dimensional transportation calculation is conducted according to the information obtained in the step 1, and neutron flux density distribution, a cell-homogenized cross section and the radial leakage item are acquired; 4, whether a characteristic value and the three-dimensional neutron flux density are convergent or not is judged, if not, iteration is continuously conducted from the step to the step 2 till the problems are convergent, and then three-dimensional neutron flux density fine distribution is obtained. According to the method, Legendre polynomial expansion unfolding is conducted on the radial leakage item in the axial direction, the calculation precision is improved, and two-dimensional spatial distribution of the axial leakage item is calculated, so that calculation is closer to true conditions.

The invention is applicable to the field of nuclear station fuel components and nuclear reactor cores, and provides a nuclear fuel rod, a nuclear fuel component and a nuclear reactor core. The nuclear fuel rod comprises two fuel pellets of different types in the axial direction thereof, i.e. a uranium dioxide fuel pellet and a gadolinium burnable poison fuel pellet; the nuclear fuel rod provided by the invention uses a type of fuel pellet as the main body, and the other type of fuel pellet is positioned at one end or two ends of the main fuel rod; the axial power error is improved, and the fuel utilization economy is improved, or the circulation length of the reactor core is prolonged through changing the neutron flux of the top end or the bottom end of the reactor core, so that the running performance of the reactor core is improved.

An apparatus for producing thermal or epithermal neutrons in a sample has a plurality of fast neutron sources positioned around a cylindrical or spherical moderator to maximize the neutron flux at a sample placed in a void at the center of the moderator.

The invention provides a liquid scintillator detector n / gamma ray screening method based on a support vector machine. The method includes the following steps that firstly, signals are collected, the signals generated by a neutron source are collected through a liquid scintillator detector, and the collected signals are transformed into pulse signals; secondly, neutrons and gamma rays are screened through a charge comparison method, the pulse signals obtained in the first step are screened through a traditional charge comparison method, and the lower limit of the screening energy of the neutrons and the gamma rays of the charge comparison method is determined; thirdly, an M matrix is obtained; fourthly, LIBSVM software is used for conducting classification forecasting on the neutrons and the gamma rays. The liquid scintillator detector n / gamma ray screening method can distinguish signals with energy below a screening threshold value in the traditional method, and therefore the accuracy of neutron flux measuring is easily improved.

The invention discloses a real-time monitoring device for the neutron flux in a fission reaction. The device is characterized in that a fast neutron conversion body (1), a fluorescent light reflection tube (3), a boron plastic flash body (2), a Cherenkov light reflection tube (5) and a Cherenkov radiation body (4) are arranged in the incident direction of particles in sequence; neutrons and gamma rays enter the boron plastic flash body (2) to interact with substances to generate e+ / e-, recoil protons and alpha particles, the e+ / e-, the recoil protons and the alpha particles are excited to generate fluorescent light, and the fluorescent light enters a first photoelectric multiplier tube (7) through reflection of the fluorescent light reflection tube (3) and is amplified through an amplifier (10) to obtain neutron and gamma information; after secondary particles enter the Cherenkov radiation body, only e+ / e- generates Cherenkov light, and the Cherenkov light is amplified through a second photoelectric multiplier tube to obtain gamma information; two signals are subjected to subtraction to obtain neutron flux information. According to the device, the n and gamma signals are judged in combination with the pulse rise time difference, so that the measurement precision of the pulsed neutron flux is further improved.

The invention relates to a method for acquiring neutron angular flux density in a nuclear fuel assembly. The method comprises the following steps: 1, calculating value of a primary function; 2, calculating integral of the primary function and angle variable; 3, establishing a wavelet unwrapping equation in the assembly; 4, processing assembly boundary conditions; 5, calculating the neutron angular flux density; 6, establishing the wavelet unwrapping equation with allowance; 7, correcting the neutron angular flux density; and 8, judging whether the neutron angular flux density meets the precision demand or not. According to the invention, argument variable of neutron angular flux density is unwrapped by using a one-dimensional primary function to reduce the calculation scale. By means of adopting the assembly boundary unwrapping and the unwrapping within the assembly respectively, the fuel assembly boundary is unwrapped in areas to avoid calculation of redundancy unwrapping coefficient. As the neutron angular flux density and the allowance are respectively processed, two-time equal coupling scales are just calculated, so that the unwrapping order is relatively and simply improved. Under the premise that approximate computational accuracy is realized, the coupling coefficient is prevented from square increase along with precision improvement, and the calculating efficiency is improved.

Example embodiments are directed to methods of producing desired isotopes in commercial nuclear reactors and associated apparatuses using instrumentation tubes conventionally found in nuclear reactor vessels to expose irradiation targets to neutron flux found in the operating nuclear reactor. Example embodiments include irradiation targets for producing radioisotopes in nuclear reactors and instrumentation tubes thereof. Example embodiments include one or more irradiation targets useable with example delivery systems that permit delivery into instrumentation tubes. Example embodiments may be sized, shaped, fabricated, and otherwise configured to successfully move through example delivery systems and conventional instrumentation tubes while producing desired isotopes.

The invention discloses a method suitable for the transportation burnup coupling calculation of a nuclear reactor. The method comprises the following steps that: 1: carrying out transportation calculation on the nucleus concentration of burnup step initiation to obtain a coarse mesh parameter and a microcosmic reaction rate; 2: carrying out burnup calculation by the microcosmic reaction rate and the nucleus concentration to obtain the nucleus concentration estimated by a burnup step end; 3: carrying out the transportation calculation by the estimated nucleus concentration to obtain the coarse mesh parameter and the microcosmic reaction rate; 4: dividing burnup CMFD (Coarse Mesh Finite Difference) substeps in a burnup step, and carrying out linear interpolation on the stored coarse mesh parameter; 5: updating the microcosmic reaction rate of the burnup CMFD substeps by the coarse mesh parameter obtained in the 4 and thin mesh neutron flux in the 3; 6: carrying out the burnup calculation by the microcosmic reaction rate on the burnup CMFD substeps to obtain the accurate nucleus concentration of the burnup step end; and 7: judging whether a burnup step number is consistent with an input value or not to judge whether calculation is finished or not. By use of the method, on a premise that extremely high accuracy is guaranteed, the step length of the burnup calculation is extremely enlarged, and calculation time in the whole service life of the nuclear reactor is shortened.

Discloses is a method for calculating single rod power of an overall reactor core. The method includes steps of 1), determining geometrical and material parameters of a target nuclear reactor according a reactor core structure thereof; establishing an SP3 equation set of step 0 or step 2 neutron angular flux torque density according to a multigroup neutron transport theory; 2), adopting structural grids to subdivide structural geometry zones with corresponding shapes, and unstructured grids to subdivide unstructured geometry zones; 3), establishing a segment SP3 method, subjecting the SP3 equation set in the step 2) to numerical discretion by adopting approximate processing modes which are mutually compatible under a structural grid and the unstructured grid, and acquiring neutron-flux density on all grids of the nuclear reactor core by utilizing iterative algorithm to solve discrete algebraic equation set; 4), adopting the neutron-flux density acquired in the step 3) to calculate the single rod power of the overall reactor core. By reducing approximation from grid subdivision, approximation from the numerical discretion and iterative calculation process and approximation from component homogenization and component power reconstitution, the high-precision single rod power of the overall reactor core can be calculated.

The invention discloses a pressurized-water nuclear reactor structure. The structure comprises an integrated reactor top, an actuating mechanism, a pressure vessel, reactor internals and a heat-insulating component, wherein the integrated reactor top is connected with a top cover of the pressure vessel, the heat-insulating component is arranged on the external surface of the pressure vessel, the reactor internals are hung to a supporting step of the pressure vessel, and the actuating mechanism is mounted on the top cover of the pressure vessel. According to the pressurized-water nuclear reactor structure, a neutron flux measurement detector, a temperature measurement detector and a water level measurement detector can be led out of a reactor through a reactor internal measurement guide structure and a pressure vessel top penetrating piece which are located in an upper closure head of the pressure vessel, so that the increase of number of reactor internal measurement tube holders on the top cover is avoided; the requirements on water level measurement can be met; the flow of a coolant entering a reactor core is distributed reasonably; the technical effect of rapidly cooling the exterior of the pressure vessel of the reactor can also be achieved by a heat insulating layer.

Radionuclides are produced with a pulsed neutron flux from a multiple repetition rate staged Z-pinch machine, the pulsed neutron flux is moderated, an activatable radionuclide precursor is exposed to the moderated pulsed neutron flux, and a corresponding radionuclide from the activatable radionuclide precursor is produced. High current pulses are passed through a target plasma of fusible material enclosed in a cylindrical liner plasma composed of a high-Z plasma to generate a magnetic field that compresses the liner plasma, and generates shock waves. The shock implodes the target plasma. The shock front propagates between an outer shock front and an axis of the target plasma so it is heated through shock dissipation and by adiabatic compression due to an imploding shock front produced in the outer liner plasma to fuse light nuclei and generate alpha particles and neutrons. Alpha particles trapped within the magnetic field further heat the target plasma.

A method and apparatus for the calibration of neutron flux monitoring devices used in a nuclear reactor core. The apparatus includes a transverse in-core probe (TIP) cable with a neutron absorber located a fixed distance apart from a TIP detector. The neutron absorber may be passed within close proximity of the neutron flux monitoring device such that a perceived drop in measured neutron flux occurs, whereupon the cable may be repositioned relative to the monitoring device to ensure that the TIP detector is within close proximity of the monitoring device for purposes of calibrating the monitoring device.