Reactive Margin Research Articles

Применение принципа Понтрягина для оптимального отключения энергетического ядерного реактора

The theory of optimal control is based on two approaches, the dynamic programming method (Bellman equation) and Pontryagin maximum principle. Pontryagin maximum principle is applied in the physics of nuclear reactors when optimizing various transient processes. The mathematical justification of this theory is based on the elements of convex analysis, which is not used by physicists and engineers, so the physical issues are less studied in scientific literature. The subject of the study is a nuclear power reactor of the WWER type. The problem of minimizing its shutdown time, bypassing the iodine pit is solved and it is possible to start up the reactor at any moment after its shutdown. Analytical and numerical methods are used. The paper considers an example of applying the maximum principle for optimal control of the process of shutting down a nuclear reactor bypassing the iodine pit. A physical mathematical model of the Pontryagin principle is formulated. The process of optimal control of reactor shutdown for large and small reactivity margins is justified and calculated. Pontryagin principle does not contain an algorithm to find an optimization process; the stages of the process must be selected based on physical considerations, but these stages must satisfy the specified principle. Based on the Pontryagin principle, the results of the study make it possible to draw up a step-by-step action plan when shutting down a WWER-type reactor with any value of the reactivity margin and its switching is possible at any time after the transition process, which avoids downtime. The proposed plan can be used both in mathematical modeling of transient processes in a reactor and in reactor control systems to improve its controllability and, consequently, to improve safety.

Reactor Physics Monitoring of a Source-Driven Subcritical System in Stationary State by Deterministic and Probabilistic Deep Neural Networks

Reactivity measurement methods, like the Amplified Source Method (ASM), link observable quantities to integral physics parameters characterizing subcritical assemblies (SCAs). These methods were mostly derived from point reactor kinetics, which assumes fundamental mode distribution. However, in SCAs, external sources cannot be neglected, leading to a nonideal response, such as the detector position dependence of measured k eff . This work investigates deterministic and probabilistic deep learning (DL) in determining k eff and kinetics/subcritical parameters using core map and foil/active detector responses as inputs, which distinguishes DL from neutronics codes. Convolutional neural networks surpassed dense neural networks with higher accuracy, while assigning a strong signature to appropriate core map features. Expansion into multi-input networks, which also process reaction rates, highlighted DL’s flexibility by accurate prediction regardless of reaction type. Uncertainty quantification of DL was done using Monte Carlo (MC) Dropout and Bayesian neural network (BNN). The results favored BNN over MC Dropout, showing greater improvement with increasing data. An assessment of ASM, applicable in a SCA at source equilibrium, showed a reactivity bias of up to −3.59%Δk/k (−4.86 $). In contrast, DL had a maximum bias of only 0.38%Δk/k (0.5 $). Underestimation by ASM represents a nonconservative scenario in criticality safety, while DL proved robust against spatial effects. This demonstrates DL’s potential in ensuring reactivity margins and a safe approach to criticality in reactor operation regimes where standard techniques can fail.

Burnable absorber element (Gd and Er) use in the WWER-type reactor to increase the refueling time

In a WWER-type pressurized water reactor, high-neutron-absorbing materials can compensate for the excess reactivity. Theoretical and computational analyses were conducted in this article to use burnable absorbers placed in a fuel rod to reduce the reactivity margin for extending refueling time for WWER-type reactors. To investigate the reduction of the reactivity margin, the fuel burnup calculation was performed without a burnable absorber and with a combination of a variety of burnable absorbers, namely natural Gadolinium (Gd) and Erbium (Er), applying a simplified GETERA program. The calculation found that the variation of the quantity of the burnable absorber (Gd, Er) inside the fuel assemblies governs the reactivity margin for the fuel burnup and increases the efficiency of Uranium fuel (UO2). The combined use of Gd and Er leads to a smoother decrease in reactor reactivity due to a smaller Er absorption cross section, which makes it possible to reduce the total mass of Gd in the fuel element and reduce the blocking effect. In the calculation, Gd concentrations employed in the computation at 1.5% and 3%, while Er concentrations were used within the range of 0.1%–0.6%.

Plutonium Production under Uranium Constraint

Production rates of fissile materials are often used to independently assess the number of nuclear warheads a state may possess. One key constraint of a plutonium-based nuclear weapons program is the availability of natural uranium, where a shortage of uranium will constrain plutonium production in the fuel cycle. Recycling of the reprocessed uranium can be used to mitigate such a shortage. Furthermore, since military reactors operate in short cycles to ensure that the plutonium is weapon-grade, it may be possible to operate them using slightly depleted uranium, provided that there are sufficient reactivity margins. Using slightly depleted or recycled uranium, the plutonium production can increase by a factor 2–5 as compared to a once-through scenario, for the same input of natural uranium. For future assessments of a state’s plutonium production, a uranium constraint should only be considered if there is clear evidence that no nuclear fuel cycle involving uranium recycling is implemented, or if evidence exists that the recycling is insufficient to mitigate the constraint.

Nuclear data uncertainty influence on the breeding ratio in sodium-cooled fast reactor systems

Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pun,γ, 238Un,γ, and 238Un,n'. The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.

Simulating a lead-cooled reactor campaign using the EUCLID/V1 code

The paper presents the results of the development of the EUCLID/V1 integrated dynamic code designed to analyze and justify the safety of fast neutron reactor facilities with a liquid-metal coolant, in terms of simulating the reactor campaign. The relevance of this study lies in the need to simulate the behavior of the core at any time during the campaign. It lets us to obtain a full dataset for subsequent simulations of the reactor dynamic conditions (including transient states or accidents). The authors have developed a fuel archive to store calculated data in HDF5 format, created a computational model editor to generate input data in the fuel archive format, and also provided an example of computing the campaign of a lead-cooled fast reactor for three core design models shown in this paper. The main array of fuel assemblies was simulated as a single unit in the first model, as three units in the second model, and in the third every single assembly was unique. In addition, the authors have shown changes in the total masses of actinides in the core, revealed that the different core models have an insignificant effect on the evolution of the total masses of actinides, and given the fuel assembly burnup values for the three core models. For the third model, the largest difference between the minimum and maximum burnup values was obtained with an almost identical average over the fuel assemblies. The reactivity margin over time for the three core models was presented. It was shown that the values and behavior of the reactivity margin during the three micro-campaigns are almost equal. From the fourth to the sixth cycle, the reactivity margin value for the third core model was lower than for the first and the second ones. Finally, the authors conclude that it is desirable to evaluate the behavior of the reactivity margin for lead-cooled fast reactor campaigns based on the detailed model of the core.

Study on an open fuel cycle of IVG.1M research reactor operating with LEU-fuel

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2–4 βeff for 20 years, and achieving a burnup of 9.9–10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

The Use of Burnable Absorbers in Vver-Type Reactors to Reduce the Fraction of the Reactivity Margin Compensated by a Liquid System During Extended Campaigns

The Use of Burnable Absorbers in Vver-Type Reactors to Reduce the Fraction of the Reactivity Margin Compensated by a Liquid System During Extended Campaigns

Comparative Assessment of the Range of Spectral Regulation of Reactivity Margin for Fuel Burnup in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

Comparative Assessment of the Range of Spectral Regulation of Reactivity Margin for Fuel Burnup in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

The Use of Burnable Poisons in VVER-type Reactors to Reduce the Fraction of the Reactivity Margin Compensated by the Liquid System During Extended Runs

The paper presents the results of a computational and theoretical analysis of the use of various burnable poisons placed in a fuel rod for the maximum reduction in the reactivity margin compensated by a liquid system based on a boron absorber for VVER-type reactors operating in extended campaigns. Various layouts of fuel rods placement in fuel assemblies and with different concentrations of burnable poisons are considered, which are considered natural gadolinium and erbium. The analysis is carried out using simplified models of fuel burnup at partial refueling.

Simulating the fuel cycle of a lead-cooled fast reactor

The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.

Dynamic Zoning and Collaborative Control of Autonomous Region of an Active Distribution Network Based on Synchronous Measurement Information

With the continuous access of distributed power to the active distribution network, the voltage overrun at key load nodes is easy to occur, which will affect the safe operation of the active distribution network. Therefore, the dynamic partition and collaborative control method of autonomous region of active distribution network based on synchronous measurement information are studied, so as to rationally dynamically partition the active distribution network, effectively coordinate the voltage control of each partition, and ensure the safe operation of distribution network. Synchronous phasor measuring device is used to collect node voltage related information of distribution network. Based on the collected information and the eigenroots of the electrical distance matrix, the number of larger eigenroots, namely the optimal dynamic partition number, is determined. Ward distance clustering algorithm is used to complete the dynamic partition of distribution network. After the completion of dynamic partition, for the nodes with the most serious voltage overlimit in each partition, the collaborative control of the voltage in each partition is completed according to the eventdriven mode. The experimental results show that the proposed method can be used to dynamically partition the autonomous regions of active distribution network and ensure that the reactive power reserve margin of each region is higher than the minimum standard. This method can effectively control the voltage of each node, and the control time is short and the active power loss is less.

Development Of Scenarios For Controlling The Fuel Campaign Of The Ivg.1M Reactor With Leu-Fuel

The paper provides calculation of various campaign scenarios of the IVG.1M Research Reactor with LEU-fuel. The schemes of replacement and transfer of the fuel are suggested. Several options are considered that include a complete and partial replacement of spent nuclear fuel with fresh one, change in the reactor reactivity margin during campaign is calculated.

The method of the reactivity margin calculation required for the IGR reactor start-up in the “Pulse” mode

The manuscript proposes a method to carry out of the Impulse Graphite Reactor (IGR reactor) start-up, which increases the reactor tests safety by reducing the number of reactivity control elements (control rods) of the IGR reactor control and protection system involved in the reactor operation.Based on the experimental data, the correlations between the parameters of the reactor power diagram and the reactivity margin required for its implementation were determined.The efficiency of the proposed method was verified in practice using the automatic power control simulation circuit and IGR reactor start-ups.

Time History of Performance Parameters of WWR-K Reactor during Gradual Replacement of the Water Reflector by a Beryllium One

The water-cooled WWR-K research reactor has been operating since 2016 using low-enrichment uranium fuel. In order to maintain high levels of generated power and reactivity margin sufficient for a 21-day operation cycle, the water-cooled core reflector of neutrons was gradually, depending on the fuel burnup, replaced by a beryllium one in order to prevent decreasing of the reactor performance. A series of calculation studies are performed with a simulation of the core region of the WWR-K reactor using an MCNP transport code including determination of neutron flux density, reactivity, efficiency of the control rods of the control and protection systems, as well as kinetic parameters critical for ensuring safe reactor operation. The calculated curves of the principal performance characteristics of the WWR-K reactor are presented and analyzed in terms of their use in scientific research and their effect on the reactor operation safety.

ВЛИЯНИЕ ЗАМЕНЫ КОНСТРУКЦИОННОЙ СТАЛИ ТВС НА ЗАПАС РЕАКТИВНОСТИ В РЕАКТОРЕ БН-600

The current program in Russia to increase the fuel consumption of fast reactors and increase its burn-out causes the transition to new structural materials, which, in turn, leads to changes in the neutron-physical characteristics of reactors. In particular, the drop in the reactivity reserve noted in the BN-600 reactor of the Beloyarsk NPP at the end of 76 operational cycles, as will be shown below, is due to the transition to a new type of shell steel with an increased content of nickel, which strongly affects the reactivity. Design support for the operation of the BN-600 and BN-800 fast reactors, as well as the experiments carried out on them, is performed by IPPE. This article presents the results of a calculated analysis of the expected changes in the reactivity reserve at the end of 76 operational cycles when replacing the shell steel in BN-600. In addition, the influence of experimental assemblies located in the core on the reactivity reserve of the BN-600 is analyzed. Analysis of calculations of the actual loading of the BN-600 reactor at 76 operational cycle using the methods of the 1st-order perturbation theory, strict perturbation theory, and the Monte Carlo method showed that a partial transition at 76 operational cycle to EK-164 shell steel leads to a decrease in the reactivity margin by 0.030±0.004 %Δk/k. Replacement of steel for the entire core will reduce the reactivity margin by ~0.12 %Δk/k, which is confirmed by Monte Carlo calculations. The calculated reactivity margin obtained at the end of 76 operational cycles for the hot state of the BN-600 reactor is in good agreement with the measured reactivity margin.

Voltage stability improvement with coordinated ULTC–STATCOM controller and VSC-HVDC in high wind penetration cases

In recent years, implementing renewable energy systems such as wind power systems into interconnected grids play a vital role in energy management and production. Besides the good aspects of this energy policy, there may be some undesirable cases if not controlled. Increasing penetration of wind generation into the electricity grid negatively affects the power system stability and the required reactive power that might cause serious voltage stability problems. In order to overcome this problem, especially in the weakest areas of the system with respect to voltage stability, it is necessary to support the reactive power reserve. In this study, a coordinated ULTC–STATCOM controller has been designed to improve the reactive power reserve and enhance voltage stability by the penetration of wind power through the VSC-HVDC wind farm system into weak area of the AC system. The novelty of the proposed coordination control than the classical coordination approaches is changing the tap position of the ULTC especially at high penetration levels and thus to maintain the reactive power reserve of STATCOM. The proposed approach is tested in a two-bus sample system and adapted to the weakest area of the IEEE 10-machine 39-bus test system that simulations are conducted in DIgSILENT PowerFactory 15.0 software. Comparative simulation results show that the proposed approach has enhanced both reactive power reserve and voltage stability margin of the test systems.

ПЕРЕВОД РЕАКТОРОВ РБМК-1000 НА РЕГЕНЕРИРОВАННОЕ ТОПЛИВО С ПОВЫШЕННЫМ СОДЕРЖАНИЕМ ЧЕТНЫХ ИЗОТОПОВ УРАНА

The structure of the Russian nuclear power industry includes reactors with different designed fuel enrichment. It is possible to mix, in certain proportions, nuclear fuel (NF) from various reactors, thus closing the nuclear fuel cycle. Reprocessed uranium is a product of radiochemical reprocessing of spent nuclear fuel (SNF) from NF with a high initial enrichment. Use of uranium-erbium fuel based on reprocessed uranium is planned for the RBMK-1000 reactor. Along with 235U and 238U, SNF contains non-fissionable ballast isotopes of uranium (232, 234, 236 U). The 232,234U isotopes have a relatively high radioactivity and the presence of these leads to an increased dose rate of ionizing radiation but, due to their small content in fuel, does not affect the neutron balance, the neutron multiplication factor, and the reactivity margin. A large presence of 236U requires additional enrichment with 235U due a greater probability of inefficient neutron absorption by the 236U nuclei. This absorption with no fission leads to a reduced neutron multiplication factor, a reduced reactivity margin in fresh fuel, and a smaller burn-up of unloaded fuel. Analyzing the effects the increased content of even uranium isotopes (IEI) has on the reactor’s neutronic performance and fuel burn-up makes it possible to determine the amount of additional 235U fuel enrichment to make up for the negative effects of 236U on the RBMK-1000 neutronic performance.

РАСЧЕТНОЕ МОДЕЛИРОВАНИЕ ДОЖИГАНИЯ ТВС ЭНЕРГОБЛОКОВ № 1 И № 2 ЛЕНИНГРАДСКОЙ АЭС ПОСЛЕ ВЫВОДА ИЗ ЭКСПЛУАТАЦИИ

The RBMK-1000 reactor operates in a continuous refueling mode. The reactor core comprises FAs across the burn-up spectrum, from fresh fuel assemblies to the most burnt-up ones. Following the reactor shutdown for decommissioning, the majority of irradiated fuel assemblies (IFA) have the potential for further use (the so-called afterburning) at operating NPP units. Most of the irradiated fuel assemblies have a burn-up fraction, which is far from the specified threshold value. The neutron multiplication factor in the cell exceeds the core average value and, therefore, the IFA loading makes it possible to increase the reactivity margin to the desired value. Reuse of spent fuel assemblies in reactors at other power units offers a number of advantages. In economic terms, afterburning is able to reduce the use of fresh fuel assemblies (FFA). In terms of radioactive waste handling, afterburning is able to reduce the number of irradiated fuel assemblies, which require long-time storage in a spent nuclear fuel repository thus reducing the IFA and radioactive waste disposal loads. JSC NIKIET has proposed the use of irradiated fuel assemblies from units 1 and 2 of Leningrad NPP shut down after decommissioning in reactors at units 3 and 4 of the same NPP.

Improvement Congestion Management in Deregulated Power System Considering Voltage Stability and Loss Minimization

First of the practical defiance's in unsystematized electrical networks is Congestion management. Two manners of percussions used in Congestion management are frill manners and cost free manners. In this paper work congestion is released by using cost free manners seeing (Flexible AC Transmission Systems) FACTS controller like SVC (Static Var compensator), STATCOM, TCSC (Thyristor controlled series Compensator), and SSSC (Static Synchronous Series Compensators) devices. Manifold-objective functions are presumed for congestion management. Small signal stability, voltage profile, and Real power loss reduction are considered objective functions in this paper. The optimal placement of FACTS controller are institute by using sensitivity founded Eigen value analysis and Continuation Power Flow (CPF). The suggested algorithm has used to define optimal placement of FACTS controller and deciphering optimal power flow (OPF) to advance voltage profile and minimize the real power losses within real and reactive power generation margin. The recital analysis has been implemented for IEEE 39 bus test system using MATLAB/PSAT (toolbox) software. Results display the suggested technique has a competency to advance the “Voltage profile, small signal stability, Loss minimization”.