Neutron Energy Distribution in a Nuclear Reactor Taking Account of the Finiteness of the Neutron Lifetime

Since the neutron lifetime is finite, an equilibrium state is not reached between neutrons and moderator nuclei. Correspondingly, an equilibrium distribution is not established. Explicit expressions for the lifetime of neutrons in a reactor are used. A single stationary nonequilibrium neutron energy distribution for the entire energy range is obtained. The qualitative behavior of the distribution obtained agrees with the measured values of this quantity. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 348–353, November, 2005.     

Main results of the first phase of the upgrade of the SM core

The first phase of the work on checking the main assumptions of the concept for upgrading the core of the SM reactor has been completed. A full-scale reactor experiment has been performed for the purpose of creating in the reactor the conditions necessary for accelerated high-dose irradiation of materials, meeting the requirements for the fast-neutron flux density, water temperature, pressure, and composition, and making it possible to install large-diameter experimental channels and apparatus for regulating the temperature and neutron spectrum. The decrease of the fuel volume and excess reactivity are compensated by a 20% increase of the uranium content in a fuel element and by replacing the corrosion-resistant steel fuel-assembly jackets with zirconium-alloy jackets. The results of the calculations and experiments performed during the first phase have shown that the objective has been achieved-the reactor can be operated efficiently with the new arrangement of the core. The objective and problems of the second phase are formulated: increase of the neutron flux density in the experimental channels by a factor of 1.5 by increasing the power density in the core, using a fuel element with a low harmful absorption of neutrons, and equalizing the power-release distribution by using a consumable absorber. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 86–92, February, 2007.     

Relative yield of delayed neutrons and half-life of their precursor nuclei with fissioning of 239Pu by 14.2–17.9 MeV neutrons

The energy dependence of the relative yield of delayed neutrons and of half-lives of their precursor nuclei with fissioning ²³⁹Pu by 14.2–17.9 MeV neutrons is measured. The data obtained are analyzed in terms of the average half-life 〈T〉 of the precursor nuclei of the delayed neutrons. It is shown that the observed growth of 〈T〉 with increasing primary neutron energy in the energy range studied is due predominately to opening of the channels for the emission fission of nuclei. Comparing the data on the average half-life of the precursor nuclei of delayed neutrons obtained in the present work with those obtained by other authors showed that the anomalously high values of 〈T〉 are due to methodological features of the experiment which are related with the properties of the reaction T(d, n)⁴He used as a neutron source on electrostatic charged-particle accelerators. The data obtained are of great value for developing instrumentational nondestructive means for determining the nuclide composition of nuclear fuel based on the detection of delayed neutrons. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 441–444, December, 2006.     

Measurement of the average neutron generation lifetime in IBR-2

The purpose of this work was to measure the lifetime of a neutron generation in an IBR-2 core as a function of its state and the core environment. The main problem was to study the possibility of decreasing the duration of a neutron pulse. Different measurement methods were used. The main one was the α-Rossi method. For comparing with experiment, the lifetime was estimated by a computational method. It was shown that the results obtained by all measuring methods used agree with one another. For the standard state of a reactor, the lifetime of neutrons in the IBR-2 core is 62 ± 2 nsec. The contribution of individual elements of the core and its surroundings to the total lifetime of a neutron generation is presented. It is noted that in experiments with part of the radiation shielding moved away there are discrepancies in the estimate of the effective fraction of the delayed neutrons of a factor of 1.5 as compared with the standard state of the reactor. No explanation has been found for such a discrepancy. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 166–172, September, 2007.     

The Void and Hole Effect in D2 Moderated Reactor

The reactivity cbange due to increase in the radius of empty hole was measured in a D₂O moderated reactor and some results differing from experiments with ZEEP were obtained. It can be concluded that the streaming in a hole is not so effective for reactivity. In measuring neutron flux in a void, a flat thermal neutron flux distribution was obtained and it has been concluded that the neutrons leaking through the empty hole or the void do not consist of thermal neutrons but fast neutrons for the most part. The experimental result of reactivity change due to the void location in the core indicates that the relation between the void location and the reactivity change is independent of the neutron flux distribution.     

Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step- ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the Safety Analysis Report.     

Fuel burnup related change of the reactivity effect in IBR-2 from 1982 to 2006

The dependence of the change of reactivity on energy production is obtained from an analysis of IBR-2 operation during the period 1982–2006. It is shown that at the start of reactor operation, aside from the pure effect of burnup, additional positive effects which are most likely associated with fuel densification and structural change of the core material operate. These effects decrease with time and go to zero. After 40000 MW·h only the effect of pure burnup remains, and from this moment the reactivity decreases linearly with coefficient k_b = −4.3·10⁻⁵%/(MW·h). A formula is obtained for calculating the coefficient of energy release at any moment of operation of the reactor. __________ Translated from Atomnaya énergiya, No. 104, No. 3, pp. 147–152, March, 2008.     

Neutron-capture therapy with ultra-cold neutrons

The status of neutron-capture therapy of malignant tumors and its problems – damage to healthy tissue as a result of neutron transport to the irradiation location and presence in the therapeutic beam of a background consisting of γ rays and fast neutrons – are presented. To solve these problems, the authors have proposed using ultracold neutrons with energy less than 10^–7 eV, whose unique capability is to undergo total reflection from the surface of a condensed substance at any angle of incidence. Numerous works have demonstrated that such neutrons can be transported along neutron guides. The cross section for inelastic scattering of neutrons by hydrogen-containing substances – water, ethyl alcohol, and biological tissue – has been measured in an IR-8 beam of ultracold and very cold neutrons. At temperature 200–300 K, the experimental data are in very good agreement with calculations, but as temperature decreases further a discrepancy appears, which could be due to the inaccuracy of the model spectra of the oscillations hydrogen-containing substances used in the calculations. The use of ultracold neutrons opens up new possibilities of neutron-capture therapy for treating malignant tumors localized in body cavities or organs.     

Use of low neutron leakage fuel loads to decrease the radiation load on a VVÉR-1000 vessel

It is shown on the basis of data obtained at Ukrainian nuclear power plants that fuel loads with low neutron leakage can be used effectively to decrease the radiation load on the reactor vessel. The characteristics of 104 fuel loads and the results of a determination of the radiation load on the vessel are analyzed to develop a criterion according to which a VVéR-1000 fuel load can be classified as a load with low neutron leakage. It is shown that the following condition can be chosen as such a criterion: the run-averaged relative power release in all protruding fuel assemblies must be less than 0.57. Different variants of the arrangement of the VVéR-1000 core are examined and analyzed. It is shown that placing burned-out fuel assemblies along the periphery of the core and decreasing the number of neutrons leaving the core do not always result in a lower neutron load on the reactor vessel. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 93–97, August, 2006.     

Developments in nondestructive reactor fuel composition analysis

Methods and equipments are described that have been developed at the Radium Institute between 1973 and 1998 for nondestructive analysis of fissile materials present in spent and unirradiated reactor fuel. The methods are based mainly on recording the neutrons and γ rays emitted by the fuel-pin assemblies for various types of reactor. Sometimes, methods are used for recording the induced neutron emission on irradiation with neutrons from isotope sources. V. G. Khlopin Radium Institute. Translated from Atomnaya énergiya, Vol. 86, No. 5, pp. 343–348, May, 1999.     

Reconstruction of metal embrittlement in a reactor vessel of the atomic icebreaker Lenin

Data from a study of radiation damage to the vessel of a reactor from the retired atomic icebreaker Lenin are used to determine the radiation embrittlement characteristics of the metal. Irradiation by a low neutron flux of 10¹⁰–10¹¹ cm⁻²sec⁻¹ at the beginning of operation is found to correspond to more intense embrittlement of the metal. Then, apparently, as harmful elements are depleted in the matrix of the metal, embrittlement is reduced until there is a change in sign relative to the standard curve obtained for neutron fluxes above 10¹³ cm⁻²sec⁻¹. It is proposed that, because of irradiation by low fluxes of neutrons in the peripheral zones of reactor vessels during some stages of operation, these zones may be damaged to a greater extent than those lying closer to the core. The irradiating neutron flux is a factor that influences the embrittlement of reactor vessel materials, so there is some interest in studying how material is damaged in the vessels of power reactors with low radiation loads which are under development. This is also needed in order to evaluate the efficacy of measures undertaken to reduce the effect of neutron irradiation on reactor vessels. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 201–205, October, 2008.     

Effect of <Superscript>3</Superscript>He and <Superscript>6</Superscript>Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of ³He and ⁶Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating ³H, ³He, and ⁶Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The ³He and ⁶Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 84–88, February, 2008.     

An analytical solution for the point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model

The point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model are solved analytically. The analytical solution is based on an expansion of the neutrons density in powers of the small parameter, the prompt neutrons generation time, into the second order differential equation in the neutron density. The relation between the time and the reactivity for reactor excursions near prompt critical is derived. Also, the neutron density and the average density of delayed neutron precursors as functions of reactivity are presented. The relations of reactivity, neutron density and temperature with time are calculated, drawn, and compared with other analytic method.     

Neutron number first-passage time distribution and the reactor time constant

The reactor time constant is described deterministically. In the present work, the random variable corresponding to the reactor time constant – the random time at which the random neutron number process in the reactor reaches a prescribed level – is studied stochastically. The statistical distribution containing the first-passage time as a thermodynamic parameter is used. The reactor time constant is related with the dynamics of the system. The neutron number first-passage time distribution is also described taking account of the system dynamics over the entire evolution time of the system. The reactor kinetics in powergeneration regimes, self-regulation due to feedback, and the behavior of a transient process with a positive reactivity perturbation are considered as examples.     

The effect of delayed neutrons on the time of establishing a stable fission chain

The stochastic process of delayed neutron multiplication is considered in limited fission chains in a reactor which is supercritical with respect to prompt neutrons. Equations are obtained for the instants of distribution of the output of delayed neutron precursors and for the mean square deviation. The effect of delayed neutrons on the time of establishment of a stable fission chain is investigated. Equations are compiled and their approximate solutions are given. It is shown that for a small reactivity and with a weak source, the average time for establishing the first stable fission chain can be reduced by the delayed neutrons by a factor of 10 or more.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 578–583, June, 1965     

Parameters optimization in a hybrid system with a gas dynamic trap based neutron source

A task of long-lived transuranic isotopes utilization is considered to be one of the urgent problems for the nuclear reactor technology. Using sub-critical hybrid systems is a possible solution of the problem. Budker Institute of Nuclear Physics SB RAS together with Nuclear Safety Institute RAS is working on a hybrid system with a neutron source based on the gas dynamic trap and sub-critical fuel blanket. This article presents the results obtained from a series of numerical experiments aimed at estimating the optimal system. Particularly, maximum neutron source emission rate has been estimated to reach 1 × 10¹⁸–2 × 10¹⁸ neutrons/s at the input parameters typical for such a system. Pb–Bi buffer zone impact on integral characteristics of fuel blanket has been considered. Decrease in amount of secondary fission neutrons as the result of buffer zone thickening has been revealed.The codes developed to conduct the investigations are also described in the article. The first one, GENESYS, is a zero-dimensional code aimed at modelling plasma processes in the gas dynamic trap. The second one, NMC (Neutron Monte-Carlo), is a Monte-Carlo particle transport code and is developed as a multipurpose tool for neutron transport calculation.     

Determination of the neutron flux and <Emphasis Type="Italic">γ</Emphasis> radiation in the core of operating and shut-down reactors using quartz glasses and element monitors

The neutron fluxes and the intensity of γ radiation are measured in 26 channels of a VVR-SM reactor and its thermal column. The fast neutron fluxes in the channels are determined using Ni, Fe, Co, Au, and Mn element monitors with different threshold energies, together with a theoretical calculation using the MCNP-4C program. The energy distribution of the neutron flux inside the fuel assembly is obtained for selected channels around the core. The flux of neutrons with energies >1 MeV is in the range (0.5–43)·10¹² cm⁻²sec⁻¹, depending on the location of the channel. A linear correlation is discovered between the induced optical absorption at the 215 nm line (E′ center) of SiO₂–BaO glass and the fast neutron flux in the channels. The γ-ray intensity in the thermal channel is estimated for the reactor during operation (∼38.4 Gy/sec) and 24 hours after it is shut down (∼24.7 Gy/sec) using the E′ centers induced in pure quartz glasses. The observed difference in the efficiency with which oxygen defects are formed during dry and wet irradiation of glass owing to the radiolysis of water must be taken into account when developing radiation technology and during the burial of radioactive waste. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 160–164, September, 2008.     

Measurement of neutron capture and fission cross sections of <Superscript>236</Superscript>U

The cross sections for radiative capture and fission for ²³⁶U in the neutron energy ranges 1–2000 eV and 1–1000 eV, respectively, have been measured. The measurements were performed on a flight baseline 5.2 m of the FAKEL linear electron accelerator at the National Research Center Kurchatov Institute using a method based on measuring the distribution of the number of particles emitted by an excited nucleus with simultaneous measurement of the energy of each particle. The spectrum of the neutrons incident on a sample was measured by detecting the reaction ¹⁰B(n, α)⁷Li. The cross sections were normalized with respect to the resonance in ²³⁶U at 5.45 eV. The data obtained are compared with the results of other experiments and evaluated data.     

Estimate of the parameters of feedback on power and the stability of IBR-2 for different average power levels

Experimental and model studies of the parameters of fast feedback on power as a function of the average power of IBR-2 have been performed. Transient power processes caused by square fluctuations of reactivity have been investigated. The changes in the parameters are estimated for average power ranging from 0.5 to 1.5 MW. The results obtained are compared with data from previous experiments performed in 1984–1996. It is noted that the influence of feedback on power decreases as the reactor operating time increases. The model of a reactor with parameters of feedback on power which correspond to one series of experiments is investigated for stability by the frequency method. It is shown that at the regular average power level 1.5 MW a reactor in a self-regulating regime (i.e., without an automatic regulator) possesses an adequate margin of stability. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 89–93, August, 2007.