RU2655014C1 - Method of determination of the fast neutrons flow - Google Patents

Abstract

FIELD: physics.

SUBSTANCE: invention relates to the field of technical physics or rather to the field of neutron detection. Method for determining the fast neutron flow comprises the steps of placing a detector in the irradiation zone, the neutron sensing element containing the nuclei ²³⁷Np, and a fast neutron flow is measured above the threshold energy of nuclear fission ²³⁷Np, while in the irradiation zone an additional detector is placed, the neutron-sensitive element containing the nuclei ²³⁸U, the fast neutron flow is measured above the threshold energy of nuclear fission ²³⁸U, and the flow of fast neutrons by threshold energy is lower than the threshold energy of fission of nuclei ²³⁷Np is determined by a linear combination of the neutron flux with energy above the threshold fission energy of the nuclei ²³⁷Np and the flux of fast neutrons with energy above the threshold energy of nuclear fission ²³⁸U.

EFFECT: increasing the accuracy of determining the neutron flux above 1/10 MeV.

1 cl, 1 dwg, 8 tbl

Description

The invention relates to the field of technical physics, and more specifically to the field of registration of neutron radiation. The invention can be most effectively used in determining the flux (flux density) of fast neutrons with energies above 0.1 MeV by fission cameras during the testing of products, devices and organisms for radiation resistance in reactors, critical assemblies and electron-nuclear installations.

A known device for measuring the neutron flux (see, for example, Fissile neutron kits. Technical description and instructions for use. Technical conditions 50 PI 2.809.040 TU. 1977, FSUE VNIIFTRI. Settlement Mendeleevo, Moscow region).

The device contains a set of radiators containing various fissionable under the influence of neutrons nuclides and recorders of fission products. Mica plates attached to the device are used as registrars.

The operation of the device is based on a visual account of the tracks created by the fission products in the recorder.

The disadvantage is the inability to measure neutron flux (flux density) during irradiation.

A known method for determining the flux (flux density) of fast neutrons is that a KNK-2-8M fission chamber with a ²³⁸ U nuclide is placed in the irradiation zone, the fission rate or the number of fission of ²³⁸ U nuclei in the chamber during the irradiation is measured and the flux is determined and neutron flux density with an energy above 1.5 MeV (see Koshelev A.S., Dovbysh L.E., Ovchinnikov M.A., Pikulina G.N., Drozdov Yu.M., Chuklyaev S.V., Pepelyshev Yu. .N. Highly sensitive detector of fast neutrons KNK-2-8M. Issues of atomic science and technology. Series: Physics of nuclear reactors. 2016. Issue 4. S. 104-115).

The method is based on measuring the fission rate of the ²³⁸ U nuclide in the KNK-2-8M chamber and determining the neutron flux (flux density) with an energy above 1.5 MeV — the threshold fission energy of ²³⁸ U nuclei by neutrons.

The disadvantage is the inability to determine the fast neutron flux with an energy above 0.1 MeV, since the threshold fission energy of ²³⁸ U nuclei is much higher than the most probable neutron spectrum energy.

Closest to the proposed technical solution is a method for determining the flux (flux density) of fast neutrons, which consists in placing a KNK-2-7M fission chamber in the irradiation zone, measuring the fission rate or the number of fission of ²³⁷ Np nuclei in the chamber during irradiation and determining flux density and neutron flux with energies above 0.55 MeV (see, for example, Koshelev A.S., Dovbysh L.E., Ovchinnikov M.A., Pikulina G.N., Drozdov Yu.M., Chuklyaev S. B. Highly sensitive detector of fast neutrons KNK-2-7M. Questions of atomic science and technology. Ser I: Physics of Nuclear Reactors 2014 Issue 3, pp 83-93)...

The method is based on measuring the fission rate of the ²³⁷ Np nuclide in the KNK-2-7M chamber and determining the neutron flux with an energy above 0.55 MeV.

The disadvantage is the inability to make a reliable estimate of the neutron flux with an energy above 0.1 MeV according to the testimony of the KNK-2-7M camera during the testing of products, devices and organisms for radiation resistance.

The essence of the proposed technical solution lies in the fact that in the method for determining the fast neutron flux, which consists in placing a detector in the irradiation zone, the neutron-sensitive element in which contains ²³⁷ Np nuclei, and measuring the fast neutron flux with an energy above the threshold nuclear fission energy of ²³⁷ Np, in the irradiation zone is placed further detector element neytronochuvstvitelny wherein the core comprises ²³⁸ U, measured the flux of fast neutrons with energy above the threshold energy fission ²³⁸ U, and the flux of fast neutrons nergiey below the threshold energy fission Np ²³⁷ define a linear combination of the flow of neutrons with energy above the threshold energy ²³⁷ Np fission neutron flux and energy above a threshold energy fission ²³⁸ U. In this approximation the flow of neutrons with energy above the threshold energy fission ²³⁷ Np, energy neutron flux above the threshold fission energy of ²³⁸ U nuclei and a neutron flux of energy below the threshold fission energy of ²³⁷ Np, a linear function is produced.

The proposed technical solution satisfies the criteria of the invention of “novelty” and “inventive step”, despite the fame of some of the features used in it, since the combination of the above features, taken in a new sequence, allows us to determine the flux of fast neutrons with an energy below the threshold nuclear fission energy of ²³⁷ Np due to the established relationship between the neutron flux of energy above the threshold nuclear fission energy of ²³⁷ Np, the neutron flux of energy above the threshold nuclear fission energy of ²³⁸ U and the flux of fast neutrons with energies below the threshold fission energy of ²³⁷ Np nuclei in reactors, critical assemblies, and in neutron sources based on reactors, critical assemblies, and electron-nuclear installations.

The following is an example of a specific implementation of the method with reference to the accompanying drawings (Fig. 1) and tables (Table 1-8.

FIG. 1 depicts the dependences of F _m on E _pores on a variety of spectra: instant fission neutrons 1, in reactors and a critical assembly with an active zone of metal uranium 2, in and near the metal active zone 3, in a reactor hall with a metal active zone 4, in solution aperiodic reactors 5, in the BIGR 6 reactor, in the reactors of nuclear power plants 7, in hydrogen-containing moderators and n-γ converters 8.

Tab. 1 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide instantaneous neutrons in the fission spectrum.

Tab. 2 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide in uranium metal reactors and critical assemblies.

Tab. 3 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide inside and near the metal core.

Tab. 4 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide in the reactor hall with a metal core (AZ).

Tab. 5 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide in aperiodic solution reactors.

Tab. 6 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with nuclide ²³⁸ U in the BIGR reactor.

Tab. 7 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining the neutron flux with energies above 0.1 MeV and the errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide in the reactors of nuclear power plants.

Tab. 8 presents the results of calculating the coefficients in a linear combination of neutron fluxes with energies above 0.55 MeV and 1.5 MeV when determining a neutron flux with energies above 0.1 MeV and errors in determining the neutron flux with energies above 0.1 MeV when irradiating detectors with a ²³⁷ Np nuclide and with ²³⁸ U nuclide in hydrogen-containing moderators and n-γ converters.

The method is as follows.

1. Detectors are placed in the irradiation zone, the neutron-sensitive element in one of which contains ²³⁷ Np nuclei, the neutron-sensitive element in the other contains ²³⁸ U nuclei.

2. Irradiate with a neutron flux.

3. Measure the neutron flux with energy above the threshold nuclear fission energy of ²³⁷ Np according to the readings of the detector with ²³⁷ Np.

4. Measure the neutron flux with energy above the threshold nuclear fission energy of ²³⁸ U according to the readings of the detector with ²³⁸ U.

5. The flux of fast neutrons with energy below the threshold nuclear fission energy of ²³⁷ Np is determined by a linear combination of the flux of fast neutrons with energy above the threshold fission energy of ²³⁷ Np and the flux of fast neutrons with energy above the threshold fission energy of ²³⁸ U.

6. In this case, the approximation of the neutron flux with energy above the threshold nuclear fission energy of ²³⁷ Np, the neutron flux with energy above the threshold nuclear fission energy of ²³⁸ U and the neutron flux with energy below the threshold fission energy of ²³⁷ Np is performed by a linear function.

If we imagine that neutron fluxes F _0.1 , F _0.55 and F _1.5 with energies above 0.1 MeV, 0.55 MeV and 1.5 MeV, respectively, in reactors, critical assemblies and other neutron sources based on reactors and critical assemblies are described by a linear function

F (E _por ) = a⋅E _por + b,

where E _pore is the threshold energy of the neutron spectrum, then the coefficients a and b determined by the least squares method are associated with the values of F _0.1 , F _0.55 and F _1.5 according to the formulas

a = -0.612F _0.1 -0.165F _0.55 + 0.778F _1.5 ;

b = 0.773F _0.1 + 0.452F _0.55 -0.225F _1.5 .

The values of a and b calculated on the set of neutron spectra (see, for example, Sevastyanov V.D., Koshelev A.S., Maslov G.N. Characteristics of neutron fields. Instantaneous fission neutron sources, 14 MeV neutron generators, research and energy Reactors, Converters of Neutron Radiation. Handbook. Edited by VD Sevastyanov. - Mendeleevo: "VNIIFTRI", 2007), are presented in table. 1-8. The same tables show the median values a _m = (a _min + a _max ) / 2 and b _m = (b _min + b _max ) / 2. Here a _min , and _max , b _min , b _max denote the minimum and maximum values of a and b on the selected set of neutron spectra, respectively.

The median neutron flux F _{m is} described by a linear function

F _m (E _por ) = a _m ⋅ E _por + b _m .

Plots of F _m versus E _pores for various sets of neutron spectra are shown in FIG. one.

энергией выше Е_пор имеет видIn the linear portion of the load characteristic, the neutron flux

energy above E _then has the form

.

.

энергией выше 0,55 МэВ и потока нейтронов энергией выше 1,5 МэВ

методом наименьших квадратов по формулеThe value of the coefficient K is determined by the results of measuring the neutron flux

with an energy above 0.55 MeV and a neutron flux with an energy above 1.5 MeV

least squares method according to the formula

,

,

Where

,

,

.

.

энергией выше 0,1 МэВ вычисляют по формулеThe neutron flux approximated by the readings of detectors with ²³⁷ Np and ²³⁸ U nuclides

energy above 0.1 MeV is calculated by the formula

.

.

от F₀₁ определяют отношениемError κ _{θ of the} deviation of the value

from F ₀₁ is determined by the ratio

.

.

вычисляют по формулеMedian value

calculated by the formula

,

,

и

- минимальное и максимальное значения κ_θ на выделенном множестве спектров нейтронов соответственно.Where

and

- the minimum and maximum values of κ _θ on the selected set of neutron spectra, respectively.

энергией выше 0,1 МэВ связывают с

и

линейной комбинациейNeutron flux

energies above 0.1 MeV are associated with

and

linear combination

,

,

.Where

.

,

и коэффициентов А и В представлены в табл. 1-8. В тех же таблицах представлены результаты вычисления относительной погрешности Δ_0,1 определения значения

по результатам измерения потоков нейтронов

и

. Относительную погрешность Δ_0,1 вычисляют для каждого спектра нейтронов по формулеValues

,

and coefficients A and B are presented in table. 1-8. The same tables show the results of calculating the relative error Δ _{0.1 of} determining the value

according to the measurement of neutron fluxes

and

. The relative error Δ _{0.1 is} calculated for each neutron spectrum by the formula

.

.

составляет:Maximum error in determining the value

is:

9% - for spectra of instant fission neutrons of nuclei,

8% - for neutron spectra in a reactor and a critical assembly with an active zone (AZ) of uranium metal,

10% - for neutron spectra in and near metallic AZ,

9% - for neutron spectra in the reactor hall with metal AZ,

17% - for neutron spectra in aperiodic solution reactors,

15% - for neutron spectra in the BIGR reactor,

20% - for neutron spectra in reactors of nuclear power plants,

20% for neutron spectra in hydrogen-containing moderators and n-γ converters.

может быть существенно уменьшена путем уточнения набора возможных в процессе испытаний спектров нейтронов.Maximum error in determining the value

can be significantly reduced by clarifying the set of possible neutron spectra during testing.

Claims (2)

1. The method for determining the fast neutron flux, which consists in the fact that a detector is placed in the irradiation zone, the neutron-sensitive element in which contains ²³⁷ Np nuclei, and the fast neutron flux is measured with an energy above the threshold nuclear fission energy of ²³⁷ Np, characterized in that it additionally placed a detector element neytronochuvstvitelny wherein the core comprises ²³⁸ U, measured the flux of fast neutrons with energy above the threshold energy fission ²³⁸ U, and fast neutron flux threshold energy below the threshold energy actually Ia nuclei Np ²³⁷ define a linear combination of the flow of neutrons with energy above the threshold energy fission ²³⁷ Np and fast neutron flux energy above the threshold energy fission ²³⁸ U. 2. The method according to p. 1, characterized in that the approximation of the neutron flux with energy above the threshold fission energy of ²³⁷ Np nuclei, the flux of neutrons with energy above the threshold fission energy of ²³⁸ U and the flux of fast neutrons with threshold energy below the threshold fission energy of ²³⁷ Np is performed by a linear function .

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