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The IBR-2M pulsed nuclear reactor operating at the JINR Laboratory of Neutron Physics delivers an intense neutron flux on the order of 1013 n/(cm2 s). By the end of the 2030s, it will outlive its useful lifetime and will have to be replaced with a new machine. The alternative neutron sources proposed so far require extensive and costly R&D efforts, and positive results cannot be guaranteed. In this paper, we present a conceptual design of the IBR-4 reactor running on plutonium-dioxide fuel which is aimed to replace the IBR‑2M machine as a neutron source. This design fully builds on proven technical solutions tested at other nuclear facilities and, therefore, requires no extensive R&D and guarantees reliable and safe operation of the proposed neutron source. Running at a 5-MW power and a pulse repetition rate of 5 Hz, IBR-4 will deliver a thermal-neutron flux of 5.8 × 1013 cm–2 s–1. When running at a 10-MW power and a pulse repetition rate of 10 Hz, the proposed source is expected to deliver a record-high thermal-neutron flux on the order of 1.1 × 1014 cm–2 s–1.

The kinetics of a pulsed booster controlled by a high-energy proton beam is examined as a promising source of powerful neutron pulses. A subcritical multiplying assembly based on plutonium dioxide with a reactivity modulator is considered as the pulsed booster. The pulsed booster intensifies neutron pulses generated in a non-multiplying target by the accelerator protons. The parameters of the neutron pulses of a booster with different lifetime of prompt neutrons and different booster multiplication (20 and 50), i.e., with multiplication coefficients 0.95 and 0.98, are calculated. The presented calculations make it possible to pick an optimal variant of the parameters of the reactivity modulator, base level of subcriticality, lifetime of prompt neutrons, and others.

The article considers the neutronics aspect of the IBR-2 reactor optimization: whether it is possible in theory to create an IBR-2-type reactor with a neutron flux in beams above the existing 0.5 × 1013 n/(cm2 s). The calculations have shown that the thermal neutron flux theoretically can be increased to (2.0−2.5) × 1013 n/(cm2 s), but only with a complete change in the reactor design: reducing the core volume, replacing the fuel type with a denser one, and changing the beam extraction system from radial to tangential. The technical implementation of these requirements is currently a challenge.

The neutron source operating at the Frank Laboratory of Neutron Physics, Joint Institute for Nuclear Research (JINR, Dubna, Russia), based on an IBR-2M pulsed reactor with a flux density of thermal neutrons of ~1013 n/(cm2 s) will exhaust its resource and be shut down by 2034. Optimizing this reactor to provide higher neutron flux density is practically impossible. Therefore, it is proposed to consider using a subcritical pulsed neutron source driven by a proton accelerator as a substitute for the IBR-2M. At present, nonmultiplying neutron sources based on high-current proton accelerators are mostly used around the world, for example, SNS or ESS with a beam power of 3 and 5 MW, respectively. In this paper, we present a theoretical comparative analysis of different categories of pulsed neutron sources driven by high-current proton beams, which can help to find the optimum solutions. It is shown that the subcritical one-zone booster with a tungsten target and a plutonium dioxide core can be a promising option. This type of source can provide an average thermal neutron flux density of more than 5.0 × 1014 n/(cm2 s) at a proton-beam power of 0.1–0.2 MW.

Neutron noise spectra in nuclear reactors are a convolution of multiple induced reactivities. Under normal IBR-2M reactor operating conditions the full noise range can reach ± 22% with respect to average power. Slow changes in average power (induced for instance by the control rods) are thereby much smaller than the neutron noise, however they are of importance in conducting the safe operation of the reactor. The problem we address is the separation of neutron noise from the slow power-changes. For this task we used an algorithm based on the basic element method (BEM) developed at Laboratory of Information Technologies – LIT JINR. The algorithm was applied to both static and dynamic states of the reactor in the 0–2 MW power range. The average processing time for one point on a x86_64 Intel Core i5-4570 Sandy Bridge processor at 3.20 GHz, was approximately 50 μ s, which allows the use of the MSPPA-6 algorithm in real time.

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ∼400 μs, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model. Keywords: Fast periodic pulsed reactors, Prompt supercritical state, MCNP6, Russian roulette

The parameters of a mathematical model of power feedback are evaluated on the basis of an analysis of transient power processes arising in the IBR-2M reactor as a result of intentional square oscillations of the reactivity during operation in a self-regulation regime. The power feedback is described by three linear difference equations and represented in the form of three aperiodic elements connected in parallel. Two components play the role of negative and one component positive feedback. The total feedback is negative, i.e., it has a stabilizing effect on the operation of the reactor. The slowest negative component of power feedback is probably due to axial swelling of the fuel as it heats up. Since the power feedback parameters change in the course of operation, their periodically repeated evaluations are one way of securing the operational safety and reliability of the reactor.

The problem of predicting the oscillations of the main thermodynamic parameters of the core in the first loop of the sodium cooling system of IBR-2M reactor is examined. Attention is focused mainly on the prediction of the temperature and sodium flow at the entry into the core as well as the thermal power. It is shown that the prediction makes it possible to reduce by a factor of 3 the influence of slow oscillations of reactivity on the power and thereby reduce the operational requirements for the automatic power stabilization system. Neural-net prediction using nonlinear autoregression nets with feedback is proposed. The results agree with experiment to within ~5%.

The pattern recognition methodologies and artificial neural networks were used widely for the IBR-2M pulsed reactor noise diagnostics. The cluster analysis allows a detailed study of the structure and fast reactivity effects of IBR-2M and nonlinear autoregressive neural network (NAR) with local feedback connection allows predicting slow reactivity effects. In this work we present results of a study on pulse energy noise dynamics and prediction of liquid sodium flow rate through the core of the IBR-2M reactor using cluster analysis and an artificial neural network.

Ensuring the safe operation of nuclear reactors is one of the most significant tasks of nuclear power engineering. The difficulty lies in the necessity of using nondestructive methods. One of such promising methods is the neutron noise diagnostics of a nuclear reactor. In this paper, the possibility of noise diagnostics by the method of positive definite Toplitz correlation matrices is studied. The advantage of this method is its purely mathematical nature; no model of the process is employed. The method is applicable whenever the process is quasi-stochastic. The results are compared with those from the conventional spectral analysis.

In this work we present the results of a study on pulse energy noise dynamics of the IBR-2M; the study employs statistical methods of time series (pulse energy) processing and hierarchical cluster analysis. It is shown that the power spectrum changes of the pulse energy fluctuations per cycle (∼11 days) has a transition region duration of ∼3 days that takes place after the reactor has reached the nominal power of 2 MW. The power noise is subsequently divided into four stable clusters, with three of which describe the noise transition region. The fourth cluster constitutes a stable structure that does not depend on noise level (amplitude of the power spectrum) or on reactor operation time. The noise transition region is formed by the vibration of the moving reflectors once the reactor has been reached.

A nonlinear model of the IBR-2M pulsed reactor dynamics relating values of variables at discreet instants of time (when power pulses appear) is developed on the basis of the MATLAB program system. The tests of the model by simulating calculated processes in the IBR-2M reactor proved the correctness of the model. A tentative estimate of the transfer coefficient for the linear part of the automatic regulator is obtained.

A reactimeter was developed for the IBR-2 periodic-pulse reactor. The reactor kinetics was described by difference equations, which relate the reactor’s parameters, corresponding to the instantaneous and preceding power pulses, and the nonlinear dependences of the energy of the power pulse and its amplitude on the reactivity. The fact that the regulated parameter of the reactor is the relative deviation of the amplitude of the power pulse from its base (prescribed average) value was taken into account. In studying transient processes, a statistically optimal filter was used to suppress the reactivity noise due to the reactor’s construction and principle of operation. The best location for inserting the filter in the block diagram of the reactimeter was chosen and the optimal smoothing coefficient determined.

The operation of the IBR-2 pulsed reactor (1987-2006) showed that a total transfer coefficient of the power decreases during the operation of the reactor, therefore each reactor cycle has its own transfer coefficients [1] and stability margins. Hence, the study of dynamics of the new reactor IBR-2M (modernized version IBR-2, since 2011) is extremely important. In this paper, the results of an experimental and a modelling study of dynamics of the IBR-2M are presented, for the reliable and safe operation of reactor. For this purpose, model dynamics of the IBR-2M has been developed, which includes experimental data on reactor feedback parameters in period 2015-2017 at a mean power of 0.5 MW. Using the model dynamics of the IBR-2M, the stability margins of the reactor in different regimes its operation (self-regulating and automatic regulating) were estimated. It shown that during the analyzed time of the operation reactor with energy production up to 1200 MW·day significant changes in dynamics of reactor has not occurred. The reactor has been working in the stability region.

The results of an experimental study of the stochastic fluctuations of the energy of IBR-2M pulses, which are used to obtain some parameters of the reactor kinetics, are presented. A pulse energy sequence was recorded at different average power levels and the distribution parameters were calculated. An ionization chamber, placed near the core, with boron was used as the neutron detector. The results obtained made it possible to evaluate the following: the average lifetime of prompt neutrons τ= (6.5 ± 0.2)·10–8 sec, the absolute power of the reactor and the source of spontaneous-fission neutrons S sp ≤ (6.72 ± 0.12)·106 sec–1. It is shown that the experimental results are close to the computational results.

The structural design of the IBR-2 periodic-pulse reactor is described. The principal stages of the project in modernizing the reactor, the technical characteristics before and after modernization, and the results of startup measures are presented.

The main parameters of IBR-2M are presented: the effective delayed-neutron fraction βeff and the promptneutron lifetime τ, calculated using the DORT two-dimensional multigroup neutron-transport compute code and the SCALE4 code with a system of multigroup nuclear constants. For a regular IBR-2M regime βeff = 0.00216 ± 0.00007, τ = (6.5 ± 0.5)·10–8 sec, the delay-neutron value γ = 0.980, the prompt-neutron decay constant in the critical state α = 3.5·104 sec–1. The calculations showed that the effective delayed-neutron fraction for IBR-2M is identical, within the error limits, to the measured value for IBR-2, the prompt-neutron lifetime is approximately 5% longer (βeff = 0.00216, τ = (6.2 ± 0.2)·10–8 sec). It is shown that βeff and τ increase somewhat as the IBR-2 core size increases in the radial direction.

Calculations have shown that in the presence of power shedding the reactivity depends strongly on the moment of shedding relative to the start of the reactor cycle, the stopping time of the reactor after power shedding, and the number of sheddings. It is shown that reactivity effects due to power shedding continue to act for 2 days after shedding, after which these effects can be neglected. The total change of the reactivity after power shedding can range from 0.12 to –0.05%Δk/k. A region of negative reactivity forms in most practical cases after power shedding. The depth of this region can reach –0.05%Δk/k, and the duration varies from several hours to several tens of hours. Knowledge of the reactivity after power shedding makes it possible to determine more accurately the consequences of different emergency situations and to predict the critical state of the reactor when the reactor is once again brought up to power.

The basic principles of measuring and analyzing nuclear-reactor noise are described. The results obtained for the IBR-2 reactor by the noise method are presented. It is shown that analysis of the noise spectra of the power and the main reactor parameters makes it possible to find deviations from normal reactor operation at the level 10 -6 ∆k/ k of the change in reactivity.

A dynamical model of total feedback is constructed and a procedure for determining the parameters of the dynamics of the slow power-feedback components is investigated. Using the quasistationary equation governing the activity balance, the general effect of power feedback, consisting of the sum of fast and slow components of the reactivity change, is identified. The form of the curves of this effect indicates that the model of dynamic power feedback in IBR-2 for slow processes consists of three components: a proportional link and two slow inertial links (fast and slow links with positive and negative transfer coefficient and time constants of several hours and several tens of hours). The parameters of the dynamics of slow action are determined. The reactivity variation curves obtained according to the model are identical to the measured curves.

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 kb = −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.

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.

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.

The reactivity effects of two types of moving reflectors are determined in the pulsed regime of the IBR-2 reactor operating at power. For this, the shape of the power pulse on fast neutrons and the power released between pulses as a function of time were measured with good statistical accuracy. A special feature of the dynamical estimate of the neutron-physical parameters is that it was obtained by analyzing the standard operating regime of the reactor. For this, to estimate the parameters of the mobile reflector the reactor was transferred into a stationary (not pulsed) subcritical operating regime in which some fast processes influencing the shape of the power pulse could not be determined. The dynamical method makes it possible to obtain a detailed description, to a high degree of accuracy, of the change in reactivity as the moving reflector rotates.

An analysis of the IBR-2 reactor power pulse shape measured over the entire dynamic range of neutron flux variation (10 4 ), i.e. from the maximum pulse power to the background power between pulses, has been carried out. Three variants of the model describing the reactor dynamics during the power pulse have been investigated. The best approximation to the experimental data has been obtained by adding to six equations describing the effect of delayed neutrons on the power pulse of two analogous ones describing the effect of the neutrons reflected from the structural elements of the reactor. It is shown that the most probable source of additional groups of neutrons may be the neutron moderators enveloping the core as well as the elements of the biological concrete shielding that are the closest to the core. These additional groups of neutrons influence essentially the formation of the power pulse.

The process of the formation of a signal by a vacuum fission chamber (VFC) under the irradiation by a short-duration pulsed neutron flux is considered for the first time. It is shown that the collection time of charge carriers in the VFC with plane–parallel electrodes does not exceed 0.4 ns at a supply voltage of >100 V and that the square of the VFC pulse width is inversely proportional to the supply voltage. The minimum detectable flux densities for fast and thermal neutrons are estimated at 5 × 1010 and 108 cm–2 s–1, respectively, at a noise current of 10–3 A for the electronic equipment.

A technique for wide-range diagnostics of the pulse form of neutrons from the ИБР-2 periodically pulsed reactor has been investigated. Organic scintillation detectors providing a wide linearity of the amplitude characteristic in the pulsed current mode and circuits controlling the detector sensitivity have been developed. A spectral sensitivity of the detectors to neutrons and γ quanta is presented. The radiation hardness of scintillators and photomultiplier tubes and the phototube stability under long-term optic loading are analyzed. An analog-to-digital wide-range recorder and its software are briefly described. The experimental results obtained using this technique with the ИБР-2 periodically pulsed reactor are given.

A model of the IBR-2 pulsed reactor of periodic operation is developed to facilitate modeling of transitional processes. The model is constructed as a module structure with the help of discrete transfer functions of the kinetics, power feedback and automatic regulator blocks taking into account nonlinear relationships. Good agreement between the calculated and experimental dependence of changes in the power for different changes in the setting reactivity is achieved.

The paper describes the construction of a gas-filled fission ionization chamber. The chamber is compensated with respect to the current of the characteristic alpha radiation and accompanying photon radiation, and the fast and supercadmium neutrons. The characteristics of the detector and the measurements of the thermal neutron flux emitted by water and cryogenic methane moderator in the IBR-2 reactor are presented. It is shown that this chamber is best for determining the temperature of neutrons in pulsed sources. 4 refs., 4 figs.

This paper presents an artificial neural network method for long-term prediction of liquid sodium flow rate through the core of the IBR-2M reactor. The nonlinear autoregressive neural network (NAR) with local feedback connection has been considered as the most appropriate tool for such a prediction. The predicted results were compared with experimental values. NAR model predicts slow changes of liquid sodium flow rate up to two days with an error less than 5%.

Neutron noise spectra in nuclear reactors are a convolution of multiple-induced reactivities. For the IBR-2 pulsed reactor (JINR-Dubna), one part is represented by the reactivities induced by the two moving reflectors, and the other part by other sources that are moderately stable. In the present study, using recordings of the mechanical noise of the two moving reflectors, their non-linear correlations into the power spectra of the reactor are extracted using statistical analysis. The remaining noise sources are moderately stable noise and can be further monitored by other automated reactor diagnoses.