Design and optimization of fuzzy-PID controller for the nuclear reactor power control

This paper introduces a fuzzy proportional-integral-derivative (fuzzy-PID) control strategy, and applies it to the nuclear reactor power control system. At the fuzzy-PID control strategy, the fuzzy logic controller (FLC) is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region and the genetic algorithm to improve the ‘extending’ precision through quadratic optimization for the membership function (MF) of the FLC. Thus the FLC tunes the gains of PID controller to adapt the model changing with the power. The fuzzy-PID has been designed and simulated to control the reactor power. The simulation results show the favorable performance of the fuzzy-PID controller.     

Reactivity balance in the IPR-R1 TRIGA reactor

Considering that the power of the IPR-R1 TRIGA reactor, located at the Nuclear Technology Development Center, Brazil, will be increased from 100 kW to 250 kW, some experiments were done in order to evaluate the magnitude of the reactivity effects associated with the reactor operation. The core excess of reactivity obtained was 1.99 $, and the shutdown margin was 1.33 $. The reactivity needed to operate the IPR-R1 reactor at 100 kW was 0.72 $, mainly due to the prompt negative temperature coefficient. A significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels. The loss of reactivity due to xenon poisoning after 8 h of operation at 100 kW was around 0.20 $, and the highest reactivity loss value caused by a void inserted in the central thimble was 0.22 $. From the results obtained, it was possible to balance all the determined reactivity losses with the reactivity excess available in the reactor, considering the present and the future reactor power operation.     

Research reactor RBT-6

Conclusions Reactor RBT-6 is simple in construction and is easily accessible for conducting experiments. The values of neutron flux in it are high for small thermal power; this together with the large duration of continuous operation ensures the possibility of conducting a wide range of experimental investigationa. Such a reactor may be recommended as a research reactor for irradiation of samples of materials up to moderate flux values (10¹⁹–10²¹ neutrons/cm²) and for conducting experiments for studying the change in the properties of materials during irradiation, which are becoming increasingly more important. Estimates show that the number of used fuel assemblies of the SM-2 reactor are sufficient for the operation of several such reactors.If necessary, the number of experimental channels in the active zone of such a reactor can be increased by increasing the number of fuel assemblies and the thermal power. Beryllium can be used as the lateral reflector. This results in a decrease of the volume of the active zone and the thermal power of the reactor, but increases its cost.Translated from Atomnaya Énergiya, Vol. 43, No. 1, pp. 3–7, July, 1977.     

Some nonlinear problems of nuclear reactor theory

In reactor criticality calculations it is necessary to take into account the relationship between the tamperature and volume of the reactor and the neutron flux. It is frequently assumed for this, that k_eff is a function of the reactor power. The method proposed here is an approximate computation of this relationship by means of some functional of the flux (this functional can be expressed by the power averaged with respect to the reactor temperature, etc.), so that its calculation by means of k_eff is a special case of the proposed method. Results are given of calculations for a system of nonlinear equations which describe the neutron transport in one-group diffusion approximation in the plane of the reactor and the heat transfer by thermal conductivity. The results are analyzed with the object of comparing the precise and approximate solutions.Translated from Atomnaya Énergiya, Vol. 16, No. 4, pp. 304–309, April, 1964     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Subcritical power reactor with irradiation by a beam of accelerated protons

Conclusions The conversion of a nuclear power plant to operation with quite deeply subcritical reactors eliminates the primary reason for the appearance of reactivity accident situations associated with the probability of the reactor being transferred into a subcritical state and runaway of this state. There is no doubt that a nuclear power plant can in principle operate on the basis of a subcritical reactor and a high-power proton accelerator. To answer the question of whether or not it is desirable to equip nuclear power plants with accelerators, it must be kept in mind that besides achieving the main goal — complete elimination of the possibility of reactivity accidents and as a consequence of such accidents emission of solid radioactive products of uranium fission with enormous consequences for ecological and economic damage — such improvements have other important consequences. These include, for example, the possibility of constructing fuel cycles on the basis of the fuel depleted with respect to fissioning isotopes (^233,235U,²³⁹Pu), which will make it possible to decrease substantially the fuel component of the cost of a nuclear power plant; the possibility of more efficient utilization of nuclear fuel by increasing significantly the interval between loadings; and, control of the power and shielding of a reactor by changing the beam current of the accelerator. All this will make it possible, in principle, even at today's level of development of reactor and accelerator technology to build a subcritical power reactor with external irradiation with a high-energy particle beam. Institute of High Energy Physics. Translated from Atomnaya énergiya, Vol. 77, No. 4, pp. 300–308, October, 1994.     

Choosing suitable type of research reactor for Mongolia

Mongolian government said that one of the sources for energy supply will be a nuclear power. To utilize nuclear energy, it is required that Mongolia has to have sufficient number of national nuclear professionals. It is totally impossible to send all nuclear professionals to study in abroad for education degrees and do provide them with on the job training abroad. One of the key approaches to educate and train nuclear professionals locally is the need for Mongolia to have our own research reactor. The chosen research reactor will be used in various studies and for educational and training purposes. It is significantly important to get public acceptance to implement the nuclear power program. We are considering that it is suitable to choose TRIGA reactor of 300-500 kW power and of 10¹²⁻¹³ n/cm² s neutron flux.     

Cascade subcritical liquid-salt reactor for burning transplutonium actinides

A cascade subcritical liquid-salt reactor designed for burning long-lived components of the radioactive wastes of the nuclear fuel cycle is examined. The cascade scheme of the reactor makes it possible to decrease by a factor of three the power of the driving accelerator as compared with conventional accelerator-blanket systems of equal power. The fuel composition of the reactor consists of 20% Np, Am, Cm, and other transplutonium elements and 80% plutonium, which are dissolved in a salt melt NaF(50%)-ZrF₄(50%). For a 10 MW proton accelerator, 1 GeV proton energy (10 mA current) and subcriticality depth 0.05, the thermal power of the reactor is 800 MW, which permits burning ∼70 kg/yr Np, Am, Cm, and other transplutonium actinides, i.e., service five VVéR type reactors of equal power. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 116–125, August, 2006.     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

A pulsing reactor

A proposal for the design of a pulsing reactor operating on light water and enriched uranium is given in this article, and the method of calculating the physical and thermal parameters of such a reactor are indicated. It is shown that when certain definite conditions are established very stable heat generation during the neutron pulses may be obtained. In the version of a pulsating reactor discussed here, the calculations indicate that it is possible to obtain over 5 Mw power with an average thermal neutron flux greater than 10¹⁴ neutrons/cm²· sec and a maximum neutron flux (during neutron pulse) of approximately 10¹⁷ neutrons/cm² · sec in the reactor core.The author takes this opportunity to express gratitude to Academician I. V. Kurchatov for his interest in this work. The author is also grateful to U. N. Zankov for his discussion of the conclusions, and to A. K. Sokolov for his participation in a number of calculations.     

Improvement of power coefficient by using burnable poison in the CANDU reactor

For stable operation of a power reactor, the power coefficient (PC) of the nuclear reactor should be less than or equal to zero. In the CANDU reactor loaded with the recovered uranium (RU) which has a uranium enrichment of ∼0.9 wt% U²³⁵, the PC is estimated to be clearly positive over a wide power range of interest, owing to the generic positive coolant temperature coefficient (CTC) and weak fuel temperature coefficient (FTC) in the CANDU reactor. In order to improve the PC of the CANDU reactor without seriously compromising the economy, introduction of the burnable poison (BP) has been proposed in this work and a physics study has been performed to find the optimal BP material and its optimal loading scheme for the CANDU reactor loaded with the CANFLEX-RU fuel. Four potential BPs (Dy, Er, Eu, and Hf) were evaluated to find the optimal BP and various loading options of the selected BP were evaluated to determine the optimal loading scheme of the BP. From the viewpoint of the achievable fuel discharge burnup, it was found that Er is evaluated to be the best BP and the BP should be loaded within the central two fuel rings because the BP loading on the inner ring is more effective for reducing the CTC. The discharge burnup of the Er-loaded CANFLEX-RU fuel was 38% higher than that of the standard natural uranium (NU) fuel. The fuel discharge burnup can be increased further if the fuel enrichment is increased. It is shown that the discharge burnup of 1.0 wt%-enriched uranium fuel is 1.7 times higher than that of the NU fuel. This study has shown that the use of the BP is feasible to render the PC of the CANDU reactor negative, even though the slight reduction of the fuel burnup is inevitable, and thus the reactor safety can be greatly improved by the use of the BP in the CANDU reactor.     

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl₂ (LSFR), sodium (SFR), and supercritical CO₂ (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO₂ power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO₂ reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Multi purpose research reactor

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research & development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.     

Self-fueling (<Superscript>233</Superscript>U) regime for a CANDU heavy-water power reactor

The possibility of using a thorium-uranium fuel cycle efficiently with a CANDU heavy-water power reactor is examined. A computational investigation of two operating regimes of the reactor is performed. It demonstrated that it is possible in practice to accumulate the required amount of ²³³U at the first stage of operation by burning a certain amount of power-grade plutonium or enriched uranium. It is shown that using only ²³³U accumulations it is possible to operate a high-capacity power reactor in a self-fueling regime for a long time (up to 20 or more years). Substantiation of the regime is based only on the CANDU power reactor technology which has been perfected in practice.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 269–275, October, 2004.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Feasibility study of a magnetic fusion production reactor

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about $1.4 billion (1982 dollars) in either case. (The direct costs are estimated at $1.1 billion.) The production cost is calculated to be $22,000/g for tritium and $260/g for plutonium of quite high purity (1%²⁴⁰Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

Formation of neutron-physical conditions for testing in the MIR reactor VVER fuel elements in nonstationary regimes with a power increase

The creation of conditions in the core of a MIR reactor that allow safe experiments modeling the parameters of VVER fuel elements with power ramping and cyclic power change as well as with an anticipated reactivity accident is examined. The construction of experimental facilities is determined. The core configuration and control-rod positions, which make it possible to attain the required parameters with the minimum reactor power and allow experiments to be performed safely, are chosen. Information about the tests performed in the reactor is presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 279–284, May, 2008.     

Simulation of reactivity transients in a miniature neutron source reactor core

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease in value due to temperature feedback effects. The trends of theoretical results were found to be similar to measured values and the peak powers agreed well with experimental results. For ramp reactivity equivalent of clean core cold excess reactivity of 4 mk (4×10⁻³ Δk/k), the predicted peak power of 100.8 kW agrees favourably with the experimental value of 100.2 kW. The measured outlet temperature of 72.6 °C is also in agreement with the calculated value of 72.9 °C for the release of the core excess reactivity. Theoretical results for the postulated accidents due to fresh fuel replacement of reactivity worth 6.71 mk and addition of incorrect thickness of Be plates resulting in 9 mk reactivity insertion were 187.23 and 254.3 kW, respectively. For these high peak powers associated with these reactivity insertions, it is expected that nucleate boiling will occur within the flow channels of the reactor core.