SYSTEM

A methodological provision system for neutron measurements performed using neutronic setups is described. The metrological characteristics of standard sources of neutrons of ﬁ ve different groups, created based on nuclear reactors and 14 MeV neutron generators, are presented. The neutron spectra are calculated by a standard method and presented in a uni ﬁ ed analytical form as a superposition of physically validated spectra

for a fi ssile detector, where A t i is the activity of the ith detector at the time t of its measurement; N n i is the number of nuclei of the target nuclide in the ith detector; λ i is the decay constant of the product of activation in the ith detector; t 0 i is the effective irradiation time of the ith detector irradiated by neutrons; t h i is the holding time of the detector; N F i is the number of tracks of fi ssion fragments in the detector of the ith fi ssile detector; and ε F is the detection sensitivity of fi ssion events. The neutron characteristics of fi elds are determined from a system of Fredholm integral equations with discrete parameters i that relate the rate Q i of the ith nuclear reaction in the activation detector irradiated by neutrons, normalized to 1 nucleus of the target nuclide in detectors, with the differential fl ux density in the neutron fi eld ϕ(E) and cross section σ i (E) for target nuclei in the ith detector: (1) where E is the energy of the neutrons, MeV.
Since there is no exact solution of the equation, because the kernel in the integrand (cross section σ i (E) of the ith nuclear reaction), as a rule, is not determined in the entire energy range of the spectrum being reconstructed, in the calculation and an a priori neutron spectrum is formed beforehand as a primary model of the reconstructed spectrum, which subsequently is deformed taking account of the agreement of the measured rate of nuclear reactions in the neutron irradiated detectors. The more physically meaningful the a priori spectrum is and the closer it is to the desired spectrum, the more accurately the spectrum can be reconstructed at the investigated point of the neutron fi eld of a neutronic setup. It follows from the (1) system of equations that the necessary conditions for determining accurately the characteristics of the measured neutron fi eld are: 1) measurement of the rate of nuclear reactions in the detectors with minimal error (no more than 1-3%); 2) use of a set of activation detectors for measuring the spectrum with energy sensitivity to neutrons covering the entire energy range of the reconstructed spectrum; and 3) use of the most reliable cross sections of nuclear reactions in reconstructing the spectrum. A suffi cient (methodological) condition for solving system (1) is the formation of a physically validated a priori spectrum. There are several known methods of forming it. For example, a Monte Carlo spectrum at the investigated point is often used as the a priori spectrum taking account of the particular design of a nuclear reactor. Although this method is promising, its implementation requires quite accurate simulation of the design of identical components and units of a nuclear reactor that form the neutron spectrum at the investigated point. One drawback of the method is its labor intensity.
In the present work, the neutron spectra were determined by the method of [2]. A priori spectra were formed in an analytical form as a superposition of physically validated neutron measurements of spectra well-studied in world practice: the spectrum of prompt fi ssion neutrons (Maxwell, Watt, and other spectra), evaporative spectra (Weisskopf spectra), moderation spectra (Fermi spectrum), Maxwellian spectrum of thermal neutrons, and monoenergetic neutrons with a Gaussian spectrum.
Indeed, fast fi ssion neutrons with energy 10 -2 -18 MeV are generated in the core as a result of a fi ssion chain reaction of 235 U nuclei. The neutrons of the Weisskopf evaporative spectra are formed as a result of inelastic interaction of fast neutrons with a fi ssion spectrum by the reactions (n, nʹ), (n, 2n) with the nuclei of the materials of the structural elements of neutronic installations as well as the materials of the objects surrounding them, for example, benches on which they are placed, fl oors, walls, and ceiling of the experimental rooms of the nuclear reactors, and so on. The neutrons of the Fermi moderation spectrum and the Maxwellian spectrum of thermal neutrons are formed as a result of multiple elastic scattering of fast neutrons by the nuclei of elements entering into the composition of the materials of the core refl ectors as well as the moderating neutrons in the media of nuclear reactors.
Monoenergetic neutrons with energy 2.5 and 14 MeV, generated by accelerated beams of deuterons in deuterium and tritium targets in the target blocks of neutron generators via the reactions D(d, n) and T(d, n), are represented in the present work in the form of the well-known gaussian distribution with standard deviation (energy resolution) σ G .
The search for the integral Spectrum F(E) was conducted in the form of a solution of an analytical expression [2]. This representation of the neutron spectrum makes it possible to reduce the systematic error considerably in the reconstruction of neutron spectrum, compare the spectra of different types of setups in terms of the contribution of individual partial spectra I. Neutron spectrum at the center of the metal core of pulsed fast reactors in the resulting spectrum, as well as in terms of the average neutron energy in the reconstructed spectra, and so on. The indicated representation of the spectrum also makes it possible to evaluate the radioactive effect of individual components of the spectrum on the investigated objects and operationally simulate individual components in the neutron spectra in the process of developing reference fi elds. The described method of calculating neutron spectra was standardized and implemented in the integral computational code KASKAD [3,4].
The generalized results of reconstruction of certain spectra of simulation reference fi elds of neutron sources based on setups certifi ed as standards in terms of the KASKAD code are presented in Table 1. The neutron fi elds are presented in fi ve groups in accordance with the scientifi c and technical problems solved using them.
The radiation resistance of electronic and optical engineering components as well as individual assemblies and components of nuclear setups of reactors being designed is studied in the group-I neutron fi elds of reactors. The irradiation is conducted in the central channels of the reactors BARS-1, -5, BIR-2, BR-1, and BR-K1 using only fast neutrons with energy >10 keV. The diameter of the central channel where the studied objects are placed does not exceed 300 mm. The average  energy of the neutrons in the spectra of the fi elds of the indicated reactors for this group is equal to 0.7-1.34 MeV. Thermal and epithermal neutrons are absent in the spectra of these fi elds.
The investigations established that in terms of structure and characteristics the spectra at the center of the cores of pulsed fast reactors are identical to those of the standard fi elds of U 235 fi ssion spectra obtained using thermal neutrons and the neutrons from spontaneous fi ssion of 252 Cf [5].
The group II neutron fi elds are located 0.7-10 m from core center along the longitudinal axis of the experimental room of the BR-1 reactor. The average energy of the neutrons in the spectra of these fi elds lies in the range 0.293-1.268 MeV. These fi elds were used to verify the codes used to calculate the neutron spectra of nuclear reactors by the Monte Carlo method. In turn, the verifi ed codes were used to calculate the characteristics of neutron fi elds in large rooms where large objects are tested. Moreover, the indicated fi elds of the BR-1 reactor with signifi cantly different spectra were used to measure the energy dependence of the sensitivity of the neutron radiometric and dosimetric apparatus. The regularities in the formation of the energy spectra of neutrons in the experimental room of a nuclear reactor were also studied in the indicated fi elds [6].
The group-III thermal and epithermal neutron fi elds of reactors have found wide application for calibrating the apparatus of the control and safety systems of nuclear reactors, determining fuel burnup in the core of power reactors, and for analytical studies, for example, to determine trace quantities of impurity elements and ultrapure materials [7].
The characteristics of the neutron fi elds behind special moderating assemblies placed ~0.6 m from core center in the BARS-1 reactor are presented in group IV. These fi elds are modeled in terms of the energy spectrum ~500 m from the core of a fast pulsed reactor for purposes of fast calibration of high-sensitivity radiometric and dosimetric apparatus used to study low-intensity neutron fi elds at signifi cant distances (>500 m) from neutronic installations.
The neutron fi elds created in thermonuclear reactions modeled using a SNEG-13 generator are presented in group V. The neutron fi elds of this group are arranged behind special moderating assemblies at a defi nite distance from the generator's target block. These fi elds of monoenergetic 15 MeV neutrons of the SNEG-13 generator 1032 mm from the target's tritium layer without an absorber and with a steel absorber simulated according to the energy spectrum the identical fi elds of a generator ~500 m from the target.
The group IV and V neutron fi elds are supposed to be used predominantly for calibrating high-sensitivity neutron apparatus of the prototype thermal nuclear reactor being developed.
In conclusion, we note that the spectra of neutron fi elds presented in this article were reconstructed with a high degree of reliability. The standard deviation of the measured and computed rate of nuclear reactions for all studied activation detectors does not exceed 2.5%.