Development of large-scale thyristor dc circuit breaker

International Nuclear Information System (INIS)

Kobayashi, S.; Tanoue, Y.; Ikegame, H.; Matushita, T.; Sato, Y.

1981-01-01

A study for developing a thyristor dc circuit breaker that is applicable to the Tokamak device for engineering feasibility is presented. The design and test of a unit circuit breaker consisting of 4kV-3kA thyristors connected 2 in series and 12 in parallel are described. And based on the results a 50kV-24kA thyristor dc circuit breaker is conceptually designed

Regenerating method for thyristor converter used in AC inverter rolling stock; Koryu inverter sharyoyo thyristor converter no kaisei hoshiki no kento

Energy Technology Data Exchange (ETDEWEB)

Inarida, S.; Ito, S.; Nakamura, K. [Hitachi, Ltd., Tokyo (Japan)

1995-09-20

A thyristor converter is superior to GTO converter from a view point of weight and efficiency. Therefore, the thyristor converter is the best answer for AC inverter rolling stock, especially for commuter and interurban trains, in the next several years. For a thyristor converter, it is very important to prevent the commutative failure even when the category voltage decreases, in order to continue regenerating operation to sustain the brake torque. In a conventional system, a series resistance is inserted between the thyristor bridge and LC filter where the DC stage voltage is decreased enough to keep the regenerating power. The conventional thyristor enables regenerating braking in the whole operational region, but result in low power factor and efficiency. This paper proposes a new regenerating method to improve the power factor and efficiency, which keep regenerating power with no series resistance and higher DC stage voltage. Simulation results show the effectiveness of the method. 4 refs., 14 figs., 1 tab.

Experimental research on the overvoltage protection of high-power thyristor switch

International Nuclear Information System (INIS)

Chang Lei; Yang Jianhua; Liu Lie

2010-01-01

The influence of resistance-capacitance absorption circuit and static equalizing resistance to thyristor switching performance is investigated, regarding both one single thyristor and two serial thyristors. A conclusion that the higher charging voltage the higher peak value of forward current, reverse current and rising rate has been achieved. To one single thyristor, the resistance-capacitance absorption circuit makes the peak value and rising rate of forward and reverse currents much higher, but lower pulse width; different resistance values (2,10) of resistance-capacitance absorption circuit make no sense to the amplitude and rising rate of forward current, and the pulse width of reverse current. However, the higher the resistance value the lower the amplitude of reverse current, which leads to a lower current rising rate. To the two serial thyristors, the current amplitudes are not precisely half that of one single thyristor, and the forward current is a little higher while the reverse one is obviously lower; the rising time of forward and reverse currents is shortened, especially for the reverse current which is only 1/5 of one single thyristor. Notably, the capacitance impact of thyristor should not be ignored. (authors)

Specific features of the switch-on gate current and different switch-on modes in silicon carbide thyristors

International Nuclear Information System (INIS)

Yurkov, S N; Mnatsakanov, T T; Levinshtein, M E; Cheng, L; Palmour, J W

2014-01-01

The specific features of the temperature and bias dependences of the switch-on gate current in SiC thyristors are examined analytically for two possible switching mechanisms. The so-called γ-mechanism, which is highly typical of the conventional Si thyristors, is characterized by very weak temperature and bias dependences. By contrast, the so-called α-mechanism, which is very characteristic of SiC thyristors, is highly sensitive to changes in temperature and bias. If the thyristor is switched on by the α-mechanism, the switch-on gate current density decreases very steeply with increasing temperature. As a result, the thyristor can lose its working capacity at elevated temperatures due to the instability against even very weak impacts. With decreasing the bias voltage U a , the gate switch-on current increases very steeply, which can make switching the thyristor on difficult. The unintentional shunting, which is apparently present in high-voltage SiC thyristors, causes the transition from the α- to the γ-mechanism at elevated temperatures and high biases. It can be supposed that introduction of a controllable technological shunting of the emitter–thin base junction allows stabilization of the temperature and bias parameters of SiC thyristors. The analytical results are confirmed by computer simulations performed in wide temperature and bias ranges for a 4H-SiC thyristor of the 18 kV class. (paper)

Power Factor Correction for Thyristor Equipment in Glass Industry ...

African Journals Online (AJOL)

Thyristor power controllers are now widely used in the glass industry for controlling furnace temperature. While offering a number of operational advantages, they operate at lagging power factors which require correction for minimum power cost. Harmonic resonance with the utility feed, however, complicate the use of ...

Comparative performance analysis of Thyristor and IGBT based ...

African Journals Online (AJOL)

The paper systematically investigates and compares the characteristics of a variable voltage fed induction motor drive for two different types of soft starters; one based on IGBT and another based on Thyristor. Experimental validation is done using ... Keywords: Variable speed electric drives, Soft starter, Thyristor, IGBT ...

Recycling flow rate control device in BWR type reactor

International Nuclear Information System (INIS)

Fujiwara, Tadashi; Koda, Yasushi

1988-01-01

Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

Simulasi Dinamika untuk Menentukan Stabilitas Sistem Tenaga Listrik Menggunakan Thyristor Controlled Braking Resistor pada Sistem IEEE 34 Node Test Feeder

Directory of Open Access Journals (Sweden)

Andi Taufiq

2012-09-01

Full Text Available Terdapat berbagai macam metode untuk meningkatkan stabilitas sistem  tenaga listrik. Salah satunya adalah dengan menggunakan metode pengereman dinamis (dynamic braking. Generator sinkron sebagai distributed generator yang digerakkan oleh mesin diesel. Pada saat terjadi gangguan pada sistem, digunakan sebuah Thyristor Controlled Braking Resistor (TCBR untuk meredam osilasi yang terjadi. Sistem yang hendak dianalisis dinamika dan stabilitasnnya adalah IEEE 34 node test feeder. Dengan sistem ini diilustrasikan karakteristik dan keefektifan TCBR untuk meredam osilasi frekuensi rendah dan mencegah terjadinya ketidakstabilan transien sistem. Dari hasil analisis diperoleh bahwa dengan adanya penambahan TCBR (Thyristor Controlled Braking Resistor maka respon transien sistem akan menjadi lebih baik. Hal ini ditunjukkan dengan adanya penurunan overshoot dan settling timenya. Dengan demikian sistem akan menuju kondisi stabil dengan lebih cepat setelah terjadi gangguan.

Automatic generation control with thyristor controlled series compensator including superconducting magnetic energy storage units

Directory of Open Access Journals (Sweden)

Saroj Padhan

2014-09-01

Full Text Available In the present work, an attempt has been made to understand the dynamic performance of Automatic Generation Control (AGC of multi-area multi-units thermal–thermal power system with the consideration of Reheat turbine, Generation Rate Constraint (GRC and Time delay. Initially, the gains of the fuzzy PID controller are optimized using Differential Evolution (DE algorithm. The superiority of DE is demonstrated by comparing the results with Genetic Algorithm (GA. After that performance of Thyristor Controlled Series Compensator (TCSC has been investigated. Further, a TCSC is placed in the tie-line and Superconducting Magnetic Energy Storage (SMES units are considered in both areas. Finally, sensitivity analysis is performed by varying the system parameters and operating load conditions from their nominal values. It is observed that the optimum gains of the proposed controller need not be reset even if the system is subjected to wide variation in loading condition and system parameters.

Computer-aided methods of determining thyristor thermal transients

International Nuclear Information System (INIS)

Lu, E.; Bronner, G.

1988-08-01

An accurate tracing of the thyristor thermal response is investigated. This paper offers several alternatives for thermal modeling and analysis by using an electrical circuit analog: topological method, convolution integral method, etc. These methods are adaptable to numerical solutions and well suited to the use of the digital computer. The thermal analysis of thyristors was performed for the 1000 MVA converter system at the Princeton Plasma Physics Laboratory. Transient thermal impedance curves for individual thyristors in a given cooling arrangement were known from measurements and from manufacturer's data. The analysis pertains to almost any loading case, and the results are obtained in a numerical or a graphical format. 6 refs., 9 figs

Test model of the fast thyristor circuit breaker, for TORE SUPRA

International Nuclear Information System (INIS)

Bareyt, B.; Leloup, C.; Rijnoudt, E.

1984-01-01

The tokamak TORE SUPRA, permits, owing to the toroidal superconducting coils and to the poloidal field system performances, long discharges (30 s and more), for a plasma current of typically 2 MA. The poloidal field system uses the magnetic energy initially stored, for the ignition and the fast rise of the plasma current, by forcing the primary current to flow through a resistor after breaking the main rectifier current by a fast thyristor circuit breaker. In order to test the technical capabilities of such a breaker system made of fast thyristors, in series and in parallel, after a single thyristor test model T1 the series arrangement was studied on a 24 thyristor test model T2 and the parallel arrangement problems, led the manufacturer CGEE Alsthom, to build a new test model T3. (author)

Thyristor crowbar system for the high current power supplies of ASDEX upgrade

International Nuclear Information System (INIS)

Kaesemann, C.-P.; Lieshout, L. van; Huart, M.; Sihler, C.

2005-01-01

The ohmic heating system and the poloidal field coils of ASDEX upgrade are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. To protect the thyristor converters against dc overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils an overvoltage protection system was required. The paper describes the motivation for-and the design and testing of the thyristor crowbar system representing the thyristor converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience of operation on the AUG coils improving the safety of the equipment

Production of fast switching power thyristors by proton irradiation

International Nuclear Information System (INIS)

Sawko, D.C.; Bartko, J.

1983-01-01

There are several techniques currently employed by various manufacturers in the fabrication of fast switching power thyristors. Gold doping and irradiation by electron beams are among the more common ones. In all cases, the fast switching capability results from a reduction of the minority carrier lifetime of the host material by the introduction of carrier traps or recombination centers. However, accompanying this beneficial reduction in switching speed is a deleterious increase in forward voltage drop which also results from the introduction of carrier traps. Methods which minimize the voltage drop increase as the switching speed is reduced are highly desirable. One such method would achieve this by introducing the traps or recombination centers into well defined narrow regions where they will be more effective in reducing the switching speed than in increasing the forward voltage drop. Because the proton range-energy relationship in materials is relatively well defined and the lifetime reducing displacements occur near the end of their ranges, the lifetime in silicon can be reduced where desired by the precise control of proton energy. Dual energy proton beams from a tandem Van de Graaff accelerator were used in the experiments to determine whether proton beam irradiations offer advantages over other techniques. This was the subject of the present work. The results indicate that this is the preferred technique for reproducibly and rapidly processing fast switching thyristors with superior characteristics. The experimental procedure is discussed and comparisons are made with electron and neutron irradiated thyristors

Silicon controlled rectifier polyphase bridge inverter commutated with gate-turn-off thyristor

Science.gov (United States)

Edwards, Dean B. (Inventor); Rippel, Wally E. (Inventor)

1986-01-01

A polyphase SCR inverter (10) having N switching poles, each comprised of two SCR switches (1A, 1B; 2A, 2B . . . NA, NB) and two diodes (D1B; D1B; D2A, D2B . . . DNA, DNB) in series opposition with saturable reactors (L1A, L1B; L2A, L2B . . . LNA, LNB) connecting the junctions between the SCR switches and diodes to an output terminal (1, 2 . . . 3) is commutated with only one GTO thyristor (16) connected between the common negative terminal of a dc source and a tap of a series inductor (14) connected to the positive terminal of the dc source. A clamp winding (22) and diode (24) are provided, as is a snubber (18) which may have its capacitance (c) sized for maximum load current divided into a plurality of capacitors (C.sub.1, C.sub.2 . . . C.sub.N), each in series with an SCR switch S.sub.1, S.sub.2 . . . S.sub.N). The total capacitance may be selected by activating selected switches as a function of load current. A resistor 28 and SCR switch 26 shunt reverse current when the load acts as a generator, such as a motor while braking.

Thyristor voltage converter in induction electric drives with microprocessor control

Energy Technology Data Exchange (ETDEWEB)

Braslavsky, I.; Zuzev, A.; Shilin, S. [Electric Drive Department, Urals State Technical University, Ekaterinburg (Russian Federation)

1997-12-31

The paper consists of some results on developed pulse model of thyristor voltage converter which is one of the most mathematically complicated unit of electric drive. The model structure and model parameter calculating method are represented. The application of the model allows to analyse stability in `locally` by the linear pulse system theory methods with talking into consideration quantise processes within the converter. Such application provides the obtaining higher accurate results comparing with the non-linear system theory approximate methods. Logarithmic frequency characteristics are used to analyse converter dynamic features and they are represented too. (orig.) 4 refs.

A new optimized self-firing mos-thyristor device

Energy Technology Data Exchange (ETDEWEB)

Breil, M.; Sanchez, J.L.; Austin, P.; Laur, J.P.

1998-12-01

In this paper, a new integrated self-firing and controlled turn-off MOS-thyristor structure is investigated. An analytical model describing the turn-off operation and parasitic latch-up has been developed, allowing to highlight and optimize the physical and geometrical parameters acting upon main electrical characteristics. The analytical model is validated by 2D simulations using PISCES. The technological fabrication process is optimized by 2D simulations using SUPREM IV. Electrical characterization results of fabricated test structures are presented. (authors) 6 refs.

High power thyristors with 5 kV blocking voltage. Volume 1: Development of high-voltage-thyristors (4.5 kV) with good dynamic properties

Science.gov (United States)

Lock, K.; Patalong, H.; Platzoeder, K.

1979-01-01

Using neutron irradiated silicon with considerably lower spread in resistivity as compared to conventionally doped silicon it was possible to produce power thyristors with breakdown voltages between 3.5 kV and 5.5 kV. The thyristor pellets have a diameter of 50 mm. Maximum average on-state currents of 600 to 800 A can be reached with these elements. The dynamic properties of the thryistors could be improved to allow standard applications up to maximum repetitive voltages of 4.5 kV.

Selective irradiation for fast switching thyristor with low forward voltage drop

International Nuclear Information System (INIS)

Brown, J.L.

1975-01-01

The switching speed of a thyristor is increased without correspondingly increasing the forward voltage drop by selectively irradiating at least portions of the PN junction between the conducting and gating portions, and 5 to 30 percent of the adjacent surface area of the conducting portions of the device. 50 to 100 percent of the surface area of the gating portions, and the peripheral portions can also be irradiated at the same time to decrease gate sensitivity and increase blocking voltage of the thyristor, respectively. Preferably, the thyristor is irradiated with electron radiation which preferably is of an intensity greater than 1 MeV and most desirably about 2 MeV. The electron irradiation is preferably to a dosage of between about 1 x 10 13 electrons/cm 2 and 1 x 10 15 electrons/cm 2 . (auth)

Improved Turn-on Characteristics of Fast High Current Thyristors

CERN Document Server

Ducimetière, L; Vossenberg, Eugène B

1999-01-01

The beam dumping system of CERN's Large Hadron Collider (LHC) is equipped with fast solid state closing switches, designed for a hold-off voltage of 30 kV and a quasi half sine wave current of 20 kA, with 3 ms rise time, a maximum di/dt of 12 kA/ms and 2 ms fall time. The design repetition rate is 20 s. The switch is composed of ten Fast High Current Thyristors (FHCT’s), which are modified symmetric 4.5 kV GTO thyristors of WESTCODE. Recent studies aiming at improving the turn-on delay, switching speed and at decreasing the switch losses, have led to test an asymmetric not fully optimised GTO thyristor of WESTCODE and an optimised device of GEC PLESSEY Semiconductor (GPS), GB. The GPS FHCT, which gave the best results, is a non irradiated device of 64 mm diameter with a hold-off voltage of 4.5 kV like the symmetric FHCT. Tests results of the GPS FHCT show a reduction in turn-on delay of 40 % and in switching losses of almost 50 % with respect to the symmetric FHCT of WESTCODE. The GPS device can sustain an i...

Stability investigations of the ASDEX feedback system with filters for reducing thyristor noise

International Nuclear Information System (INIS)

Crisanti, F.; Schneider, F.

1983-06-01

A computer program for analysing the absolute and relative stabilities of any complex system by the root-locus method was developed. It is used to reanalyse the present horizontal position feed-back control in the ASDEX tokamak and to select the optimum parameters for this system with RCL filters for reducing thyristor noise. (orig.)

Results of test operation of the thyristor power supply for the BU-70 proton synchrotron ring electromagnet

International Nuclear Information System (INIS)

Vasil'ev, S.N.; Gusev, G.I.; Dan'shin, V.P.; Eliseenko, A.I.

1985-01-01

The thyristor converter put into operation in the course of reconstruction of the power supply of the BU-70 synchrotron ring electromagnet is described. The converter is designed for maximum current I=5 kA and voltage 2x10 kV and consists of four three-phase bridge rectifiers connected in parallel in the equivalent twelve-phase circuit. The total number of thyristors - 2304. The thyristor power supply operated over 7000 h. The conclusion is drawn that the use of thyristors allows one to sharply reduce the time of system tuning for nominal operation mode, reduce water consumption for convertercooling and considerably improve the working conditions ofthe maintenance personnel

Asymmetric quantum well broadband thyristor laser

Science.gov (United States)

Liu, Zhen; Wang, Jiaqi; Yu, Hongyan; Zhou, Xuliang; Chen, Weixi; Li, Zhaosong; Wang, Wei; Ding, Ying; Pan, Jiaoqing

2017-11-01

A broadband thyristor laser based on InGaAs/GaAs asymmetric quantum well (AQW) is fabricated by metal organic chemical vapor deposition (MOCVD). The 3-μm-wide Fabry-Perot (FP) ridge-waveguide laser shows an S-shape I-V characteristic and exhibits a flat-topped broadband optical spectrum coverage of ~27 nm (Δ-10 dB) at a center wavelength of ~1090 nm with a total output power of 137 mW under pulsed operation. The AQW structure was carefully designed to establish multiple energy states within, in order to broaden the gain spectrum. An obvious blue shift emission, which is not generally acquired in QW laser diodes, is observed in the broadening process of the optical spectrum as the injection current increases. This blue shift spectrum broadening is considered to result from the prominent band-filling effect enhanced by the multiple energy states of the AQW structure, as well as the optical feedback effect contributed by the thyristor laser structure. Project supported by the National Natural Science Foundation of China (Nos. 61604144, 61504137). Zhen Liu and Jiaqi Wang contributed equally to this work.

Improved control system of the thyristor flicker suppressor for the KEK 12-GeV PS

International Nuclear Information System (INIS)

Matsumoto, S.; Baba, H.; Mikawa, K.; Sato, H.; Sueno, T.

1983-01-01

Thyristor control system of the 20 MVar flicker suppressor has been improved essentially. The previous feed forward (FF) loop with each single phase reactive current detector of the MR magnet power supply was exchanged to the present by both FF- and NFB-loops. The FF-loops consists of a three phase reactive power detector of the MPS and a forcing pattern generator on the fast but steady line voltage flicker, sag and surge. The NFB-loops control by the slow parts of the flicker and the unbalanced line voltages. These detectors of the reactive power, the voltage flicker and the unbalance have been developed. Sampled voltage flicker data with 12 bit ADC are processed by Z-80A micro computer system and the forcing pattern is generated by the system through 12 bit DAC into the loop. A typical voltage flicker including sag and surge has been reduced within + or - 1.5%, about 1/3 compared to the previous, at 66 kV primary line

A new inverter topology using GTO commutation. [Gate Turn Off thyristor

Science.gov (United States)

Rippel, W. E.

1983-01-01

A new N-phase, forced commutated bridge inverter topology has been developed wherein a single Gate Turn Off Thyristor (GTO) is used to commutate each of 2N main Thyristors (SCRs). Since, for most applications, the primary loss mechanism is the SCR forward drop, very high efficiencies are possible. Compared with conventional pure SCR and pure GTO inverters, cost per kW is lower - in the former case due to the large cost differential between GTOs and SCRs. Other advantages of the new inverter include high power density, low switching losses and stresses, modulation flexibility and amenability to high voltage and high frequency operation.

Optical triggering of 4H-SiC thyristors (18 kV class) to high currents in purely inductive load circuit

International Nuclear Information System (INIS)

Rumyantsev, S L; Levinshtein, M E; Saxena, T; Shur, M S; Cheng, L; Palmour, J W; Agarwal, A

2014-01-01

Optical switch-on of a very high voltage (18 kV class) 4H-SiC thyristor with an amplification step (pilot thyristor) to the current I max  = 1225 A is demonstrated using a purely inductive load and a calibrated air transformer. Increasing the inductance of the transformer primary winding slows down the turn on process. However, the inductance has little effect during the initial stage of the switch-on process when the voltage drop on the thyristor and its internal resistance is high. The results show that a further switch-on current increase can be only achieved by introducing additional amplification steps in the pilot thyristor. (paper)

Method for calculation of the dynamic behaviour of permanent-magnet synchronous machines controlled by a thyristor-based softstarter; Berechnungsmethode zur Bestimmung des dynamischen Verhaltens von permanentmagneterregten Synchronmaschinen am Drehstromsteller

Energy Technology Data Exchange (ETDEWEB)

Benecke, Marcel; Lindemann, Andreas [Magdeburg Univ. (Germany). Inst. fuer Elektrische Energiesysteme; Griepentrog, Gerd [Siemens AG (Germany). Corporate Technology CT T DE

2011-07-01

Permanentmagnet synchronous machines (PMSM) show lower losses compared to induction machines (IM) due to the lack of rotor losses which are typical only for IM. Hence, for many applications with a fixed rotational speed such as fans or pumps the PMSM would be a good choice in case the PMSM is directly connected to the grid without any converter in between at nominal speed. This would be an optimal solution with respect to efficiency and losses. Because the PMSM is not able to speed up when directly connected to the grid, it has normally to be started using a frequency converter or it has to be equipped with a start-up short-circuit cage which would develop an asynchronous torque. A cost effective alternative is to use AC voltage controller or softstarter based on thyristors to speed up the PMSM and bridge the AC voltage controller once the PMSM has reached nominal speed. This paper presents a calculation method to predict the dynamic behaviour of a PMSM when fed by a thyristor-based AC voltage controller. (orig.)

Mathematical Model of Thyristor Inverter Including a Series-parallel Resonant Circuit

OpenAIRE

Miroslaw Luft; Elzbieta Szychta

2008-01-01

The article presents a mathematical model of thyristor inverter including a series-parallel resonant circuit with theaid of state variable method. Maple procedures are used to compute current and voltage waveforms in the inverter.

Mathematical model of thyristor inverter including a series-parallel resonant circuit

OpenAIRE

Luft, M.; Szychta, E.

2008-01-01

The article presents a mathematical model of thyristor inverter including a series-parallel resonant circuit with the aid of state variable method. Maple procedures are used to compute current and voltage waveforms in the inverter.

Reactive Power Control in Eight Bus System Using FC-TCR

Directory of Open Access Journals (Sweden)

Thangavelu Vijayakumar

2010-02-01

Full Text Available This paper deals with the simulation of eight bus system having fixed capacitor and thyristor controlled reactor. The system is modeled and simulated using MATLAB.The simulation results are presented. The power and control circuits are simulated. The current drawn by the TCR varies with the variation in the firing angle. The simulation results are compared with the theoretical results.

Mathematical Model of Thyristor Inverter Including a Series-parallel Resonant Circuit

Directory of Open Access Journals (Sweden)

Miroslaw Luft

2008-01-01

Full Text Available The article presents a mathematical model of thyristor inverter including a series-parallel resonant circuit with theaid of state variable method. Maple procedures are used to compute current and voltage waveforms in the inverter.

Thyristor based short circuit current injection in isolated grids

OpenAIRE

Hoff, Bjarte; Sharma, Pawan; Østrem, Trond

2017-01-01

This paper proposes a thyristor based short circuit current injector for providing short circuit current in isolated and weak grids, where sufficient fault current to trigger circuit breakers may not be available. This will allow the use of conventional miniature circuit breakers, which requires high fault current for instantaneous tripping. The method has been validated through experiments.

Protection relay of phase-shifting device with thyristor switch for high voltage power transmission lines

Science.gov (United States)

Lachugin, V. F.; Panfilov, D. I.; Akhmetov, I. M.; Astashev, M. G.; Shevelev, A. V.

2014-12-01

Problems of functioning of differential current protection systems of phase shifting devices (PSD) with mechanically changed coefficient of transformation of shunt transformer are analyzed. Requirements for devices of protection of PSD with thyristor switch are formulated. Based on use of nonlinear models of series-wound and shunt transformers of PSD modes of operation of major protection during PSD, switching to zero load operation and to operation under load and during short circuit operation were studied for testing PSD with failures. Use of the principle of duplicating by devices of differential current protection (with realization of functions of breaking) of failures of separate pares of PSD with thyristor switch was substantiated. To ensure protection sensitivity to the shunt transformer winding short circuit, in particular, to a short circuit that is not implemented in the current differential protection for PSD with mechanical switch, the differential current protection reacting to the amount of primary ampere-turns of high-voltage and low-voltage winding of this transformer was designed. Studies have shown that the use of differential current cutoff instead of overcurrent protection for the shunt transformer wndings allows one to provide the sensitivity during thyristor failure with the formation of a short circuit. The results of simulation mode for the PSD with switch thyristor designed to be installed as switching point of Voskhod-Tatarskaya-Barabinsk 220 kV transmission line point out the efficiency of the developed solutions that ensure reliable functioning of the PSD.

Correlator receiver architecture with PnpN optical thyristor operating as optical hard-limiter

Science.gov (United States)

Kang, Tae-Gu; Ho Lee, Su; Park, Soonchul

2011-07-01

We propose novel correlator receiver architecture with a PnpN optical thyristor operating as optical hard-limiter, and demonstrate a multiple-access interference rejection of the proposed correlator receiver. The proposed correlator receiver is composed of the 1×2 splitter, optical delay line, 2×1 combiner, and fabricated PnpN optical thyristor. The proposed correlator receiver enhances the system performance because it excludes some combinations of multiple-access interference patterns from causing errors as in optical code-division multiple access systems with conventional optical receiver shown in all previous works. It is found that the proposed correlator receiver can fully reject the interference signals generated by decoding processing and multiple access for two simultaneous users.

Electronic control system for irradiation probes

International Nuclear Information System (INIS)

Gluza, E.; Neumann, J.; Zahalka, F.

1980-01-01

The EROS-78 system for the supply and power control of six heating sections of the irradiation probe of the CHOUCA type placed in the reactor vessel is described. The system allows temperature control at the location of the heat sensor with an accuracy of +-1% of the rated value within the region of 100 to 1000 degC. The equipment is provided with its own quartz controlled clock. The temperature is picked up by a chromel-alumel jacket thermocouple. The power input of the heating elements is thyristor controlled. (J.B.)

A new self-firing MOS-thyristor device: optimization of the turn-off performance and experimental results

Energy Technology Data Exchange (ETDEWEB)

Breil, M.; Sanchez, J.L.; Austin, P.; Laur, J.P.

1998-12-01

In this paper, a new integrated self-firing and controlled turn-off MOS-thyristor structure is investigated. An analytical model describing the turn-off operation and parasitic latch-up has been developed, allowing to highlight and optimize the physical and geometrical parameters acting upon main electrical characteristics. The analytical model is validated by 2D simulations using PISCES. The technological fabrication process is optimized by 2D simulations using SUPREM IV. Electrical characterization results of fabricated test structures are presented. (authors) 6 refs.

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

Proposal for the Award of a Contract for the Supply of the Power Units for LHC Thyristor Power Converters

CERN Document Server

2003-01-01

This document concerns the award of a contract for the supply of 12 power units of thyristor power converters rated from 365 to 770 kW for the LHC. Following a market survey carried out among 98 firms in nineteen Member States, a call for tenders (IT-3003/SL/LHC) was sent on 25 October 2002 to six firms in four Member States. By the closing date, CERN had received five tenders from five firms in four Member States. The Finance Committee is invited to agree to the negotiation of a contract with OCEM (IT), the lowest bidder, for the supply of 12 power units of thyristor power converters for a total amount of 981 484 Swiss francs not subject to revision, with options for three additional units of the power part of thyristor power converters, for an additional amount of 249 681 Swiss francs, subject to revision for inflation from 1 August 2006, bringing the total amount to 1 231 165 Swiss francs. The firm has indicated the following distribution by country of the contract value covered by this adjudication propos...

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

The TCSC (Thyristor Controlled Series Compensator) in a multi machine electric power system: the effects on the synchronizing power; O TCSC em um sistema de energia eletrica multimaquinas: os efeitos sobre a potencia sincronizante

Energy Technology Data Exchange (ETDEWEB)

Silva, J.C.; Colvara, L.D. [Universidade Estadual Paulista (FEIS/UNESP), Ilha Solteira, SP (Brazil). Dept. de Engenharia Eletrica], Emails: jadiel_silva@hotmail.com, laurence@dee.feis.unesp.br

2009-07-01

The problem of stability of electric power systems, from the standpoint of ability to sync and how FACTS (Flexible Alternating Current Transmission Systems) devices affect this ability in particular the TSCS (Thyristor Controlled Series Compensator) inserted into an environment multi machine, is addressed. The effects of this device on the power synchronizing are considered through analysis of the matrix admittance of the bar, focusing on the transfer admittances between machines.

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Reactor-core-reactivity control device

International Nuclear Information System (INIS)

Miura, Teruo; Sakuranaga, Tomonobu.

1983-01-01

Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

A thyristor breaker of 1.5 109 V.A. for the poloidal field system of TORUS SUPRA

International Nuclear Information System (INIS)

Bareyt, B.; Leloup, C.; Rijnoudt, E.

1980-09-01

The poloidal field system of Torus Supra has an inductive storage of approximately 38 MJ, which has conducted to research the best solution for a D.C. breaker (55 kA., 27 kV). A solid-state breaker has been chosen. The working principles have been tested on a small size model. The final circuit breaker will contain a large number of thyristors in series and in parallel; the critical problem lies in the series arrangement. A test unit for full tension has been constructed. In this unit the thyristors are submitted to the maximum current as well as to the maximum voltage. The surges measured during tests are not higher than the calculated values. A synthetic circuit of the poloidal field system has been used for test under the final working conditions

Reactor power control device

International Nuclear Information System (INIS)

Imaruoka, Hiromitsu.

1994-01-01

A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

Reactor control device

International Nuclear Information System (INIS)

Araki, Takao; Inoue, Toyokazu.

1981-01-01

Purpose: To protect the reactor floor by alleviating the shock imparted to the reactor floor by a dropped control rod when a wire rope accidentally breaks. Constitution: A control rod is hung by wire rope from a control rod drive, and shock absorbers are mounted at the upper and lower portions of the control rod. The outer diameter of the upper shock absorber is made larger than the inner diameter of a control rod inserting hole formed in the reactor core. If the control rod drops, the upper absorber is stopped at the upper tapered portion of the inserting hole. Thus, the dropping energy of the control rod can be sufficiently absorbed by the upper and lower shock absorbers. (Kamimura, M.)

ac power control in the Core Flow Test Loop

International Nuclear Information System (INIS)

McDonald, D.W.

1980-01-01

This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

Nuclear reactor control column

International Nuclear Information System (INIS)

Bachovchin, D.M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

Operational experience with reactive power control methods optimized for tokamak power supplies

International Nuclear Information System (INIS)

Sihler, C.; Huart, M.; Kaesemann, C.-P.; Streibl, B.

2003-01-01

The power and energy of the ASDEX Upgrade (AUG) tokamak are provided by two separate 10.5 kV, 110-85 Hz networks based on the flywheel generators EZ3-EZ4 in addition to the generator EZ2 dedicated to the toroidal field coil. The 10.5 kV networks supply the thyristor converters allowing fast control of the DC currents in the AUG poloidal field coils. Two methods for improving the load power factor in the present experimental campaign of AUG have been investigated, namely the control of the phase-to-neutral voltage in thyristor converters fitted with neutral thyristors, such as the new 145 MVA modular thyristor converter system (Group 6), and reactive power control achieved by means of static VAr compensators (SVC). The paper shows that reliable compensation up to 90 MVAr was regularly achieved and that electrical transients in SVC modules can be kept at an acceptable level. The paper will discuss the results from the reactive power reduction by SVC and neutral thyristor control and draw a comparative conclusion

Elements on reactor control

International Nuclear Information System (INIS)

Bruna, G.B.

1998-01-01

In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

Application of superconducting coils to VAR control in electric power systems: a proposal

International Nuclear Information System (INIS)

Boenig, H.J.; Hassenzahl, W.V.

1979-11-01

During the last eight years, static VAR-control systems with thyristor-controlled, room-temperature reactors have been used in electrical systems for voltage control and system stabilization. In this proposal, we describe a new static VAR-control system that uses an asymmetrically controlled Graetz bridge and a superconducting dc coil. Preliminary studies indicate that the proposed system will have lower overall losses and that its capital cost and electrical characteristics are comparable to those of a conventional system. Three- and four-year programs for developing the electronic circuitry and superconducting coils for VAR control, culminating in the installation and testing of an approx. 40-MVAR system, are proposed

Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

Reactor control device

International Nuclear Information System (INIS)

Kameda, Akiyuki.

1979-01-01

Purpose: To enable three types of controls, that is, level control, scram control and excess reactivity control required for a reactor by a same mechanism by feeding neutron absorber liquid and pressure control gas to several blind pipes provided in the reactor core. Constitution: A plurality of blind pipes are disposed spaced apart in a reactor core and connected by way of injection pipes to a neutron absorber liquid tank. A pressure regulator is connected to the blind pipes, to which pressure control gas is supplied. The neutron absorber liquid used herein consists of sodium, potassium or their alloy, or mercury as a basic substance incorporated with one or more selected from boron, tantalum, rhenium, europium or their compounds. The level control, scram control and excess reactivity control can be attained by moderating the pressure changes in the pressure control gas or by regulating the fluctuation in the liquid level. (Horiughi, T.)

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

Fast-acting nuclear reactor control device

International Nuclear Information System (INIS)

Kotlyar, O.M.; West, P.B.

1993-01-01

A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

Reactor power control device

International Nuclear Information System (INIS)

Kobayashi, Akira.

1980-01-01

Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

To the issue of increasing efficiency of VAR compensation

Directory of Open Access Journals (Sweden)

Світлана Костянтинівна Поднебенна

2015-11-01

Full Text Available This article describes the features of VAR compensation of variable loads. One of the most common non-symmetric non-linear power consumers are welding power sources. Time and duration of the work of these sources vary randomly. To compensate the consumption of reactive power on the basis of consumption data from the three-phase electricity meters is inefficient. Compensation devices power should be calculated taking into account the asymmetrical consumption/generation of reactive power per phase and changing consumption patterns. Thyristor-controlled reactor, thyristor-switched capacitors, hybrid VAR compensators, active compensators (STATCOMs, «dynamic capacitors» can be used as the VAR compensation devices. Thyristor-controlled reactors can provide smooth regulation of reactive power, but they have high weight and size parameters and are additional sources of higher harmonics. Thyristor-switched capacitors provide stepwise adjustment of reactive power and are subject to the current higher harmonics. Hybrid VAR compensators make it possible to isolate capacitors from the higher harmonics and ensure smooth regulation, which is achieved by active filter introduction to the reactive power compensation devices based on thyristor-switched capacitors. However, this increases the cost of a compensator and complicates its control system. STATCOMs provide smooth regulation of reactive power, but they are too expensive.Perspective direction in the development of effective VAR compensation devices is «dynamic capacitor». As a result of a feasibility study the prospects for further research of electrical grids power efficiency through the development of effective devices for VAR compensation have been established

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Reactor core control device

International Nuclear Information System (INIS)

Sano, Hiroki

1998-01-01

The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

Digitally controlled twelve-pulse firing generator

International Nuclear Information System (INIS)

Berde, D.; Ferrara, A.A.

1981-01-01

Control System Studies for the Tokamak Fusion Test Reactor (TFTR) indicate that accurate thyristor firing in the AC-to-DC conversion system is required in order to achieve good regulation of the various field currents. Rapid update and exact firing angle control are required to avoid instabilities, large eddy currents, or parasitic oscillations. The Prototype Firing Generator was designed to satisfy these requirements. To achieve the required /plus or minus/0.77/degree/firing accuracy, a three-phase-locked loop reference was designed; otherwise, the Firing Generator employs digital circuitry. The unit, housed in a standard CAMAC crate, operates under microcomputer control. Functions are performed under program control, which resides in nonvolatile read-only memory. Communication with CICADA control system is provided via an 11-bit parallel interface

Volatile and Nonvolatile Characteristics of Asymmetric Dual-Gate Thyristor RAM with Vertical Structure.

Science.gov (United States)

Kim, Hyun-Min; Kwon, Dae Woong; Kim, Sihyun; Lee, Kitae; Lee, Junil; Park, Euyhwan; Lee, Ryoongbin; Kim, Hyungjin; Kim, Sangwan; Park, Byung-Gook

2018-09-01

In this paper, the volatile and nonvolatile characteristics of asymmetric dual-gate thyristor random access memory (TRAM) are investigated using the technology of a computer-aided design (TCAD) simulation. Owing to the use of two independent gates having different gate dielectric layers, volatile and nonvolatile memory functions can be realized in a single device. The first gate with a silicon oxide layer controls the one-transistor dynamic random access memory (1T-DRAM) characteristics of the device. From the simulation results, a rapid write speed (107) can be achieved. The second gate, whose dielectric material is composed of oxide/nitride/oxide (O/N/O) layers, is used to implement the nonvolatile property by trapping charges in the nitride layer. In addition, this offers an advantage when processing the 3D-stack memory application, as the device has a vertical channel structure with polycrystalline silicon.

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Digital control of research reactors

International Nuclear Information System (INIS)

Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

1991-01-01

Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

Thyristor current-pulse generator for betatron electromagnet with independent low-voltage supply

International Nuclear Information System (INIS)

Baginskii, B.A.; Makarevich, V.N.; Shtein, M.M.

1989-01-01

A thyristor generator is described that produces unipolar current pulses in the winding of a betatron electromagnet. The voltage on the electro-magnet is increased and the shape of the current pulses is improved by use of an intermediate inductive storage device. The current pulses have a duration of 11 msec, an amplitude of 190 A, and a repetition frequency of 50 Hz. The maximum magnetic-field energy is 450 J, the voltage on the electromagnet winding is 1.5 kV, and the supply voltage is 27 V

Realization of a Temperature Based Automatic Controlled Domestic Electric Boiling System

OpenAIRE

Shengqi Yu; Jinwei Zhao

2017-01-01

This paper presents a kind of analog circuit based temperature control system, which is mainly composed by threshold control signal circuit, synchronization signal circuit and trigger pulse circuit. Firstly, the temperature feedback signal function is realized by temperature sensor TS503F3950E. Secondly, the main control circuit forms the cycle controlled pulse signal to control the thyristor switching model. Finally two reverse paralleled thyristors regulate the output p...

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

Control of WWER-440 nuclear reactor

International Nuclear Information System (INIS)

Wagner, K.; Drab, F.; Grof, V.

1978-01-01

The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Reactor power control device

International Nuclear Information System (INIS)

Doi, Kazuyori.

1981-01-01

Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

4H-SiC gate turn-off (GTO) thyristor development

Energy Technology Data Exchange (ETDEWEB)

Casady, J.B.; Agarwal, A.K.; Rowland, L.B.; Siergiej, R.R.; Seshadri, S.; Mani, S.; Sanger, P.A.; Brandt, C.D. [Northrop Grumman Sci. and Technol. Center, Pittsburgh, PA (United States); Barrows, J.; Piccone, D. [Silicon Power Corp., Malvern, PA (United States)

1998-08-01

4H-SiC inverted, asymmetrical gate turn-off thyristors (GTOs) were fabricated and characterized over an ambient temperature range of 25 C to 390 C. Device performance was evaluated with respect to forward drop, current density, and blocking voltage. At room temperature, forward blocking voltages of up to 1000 V were achieved in smaller area devices (6.5 x 10{sup -4} cm{sup 2} active area) while larger area devices (3.63 x 10{sup -3} cm{sup 2} active area) could block up to 700 V. Reverse blocking was approximately 50 V for these asymmetrical devices. Current densities were evaluated up to 3500 A/cm{sup 2}, with the forward voltage drop strongly affected by temperature and anode contact resistance. (orig.) 11 refs.

Reactor control rod supporting structure

International Nuclear Information System (INIS)

Akimoto, Tokuzo; Miyata, Hiroshi.

1984-01-01

Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

Reactor control device

International Nuclear Information System (INIS)

Kinoshita, Mitsuo.

1991-01-01

Heretofore, since the aimed value of a reactor power has been determined only based on the deviation between a temperature variation coefficient and the aimed value, it involves a problem that a region for finely determining a control constant is complicated. In view of the above, in the present invention, an approximate value of the aimed reactor power value is determined based on the aimed value for the temperature variation coefficient, then a compensation value for the aimed power value is determined based on the deviation between the temperature variation coefficient and the aimed value and, further, the aimed power value is determined based on the approximate value and the compensation value. Control elements are automatically operated so that the power follows the aimed value after determining the aimed value. Then, since the aimed reactor power value is controlled finely so that the responsiveness of the temperature variation coefficient is satisfactory and the temperature variation coefficient agrees with the aimed value, the stability for the control of temperature variation coefficient is satisfactory. That is, high performance control is enabled by a simple control algorithm to reduce the number of the steps for the design and the device control. (N.H.)

reactor power control using fuzzy logic

International Nuclear Information System (INIS)

Ahmed, A.E.E.

2001-01-01

power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

Study and analysis of failure modes of the electrolytic capacitors and thyristors, applied to the protection system of the LHC (Large Hadron Collider); Etude et analyse des modes de defaillances des condensateurs electrolytiques a l'aluminium et des thyristors: appliquees au systeme de protection du LHC (Large Hadron Collider)

Energy Technology Data Exchange (ETDEWEB)

Perisse, F

2003-07-15

The study presented in this thesis is a contribution about the analysis of failures modes of electrolytic capacitors and thyristors. The studied components are main elements of the protection system of the superconductive magnets of the LHC. The study of the ageing of the electrolytic capacitors has shown that their reliability is strongly related to their technological characteristic. Evolution of their principal indicator of ageing (ESR) can be modeled according to different laws chosen according to their running mode. It appears that the prediction of failure of these components other than that due to wear can be only statistical taking into account the many causes of failure involving various modes of failure. In order to be able to evaluate influence of the ageing of the electrolytic capacitors on a system, simple models taking into account this parameters as well as the effective temperature of the component are proposed. An acceptable precision taking into account the simplicity of the models is obtained. The study of the thyristors has shown that these components have little drift of parameters in static ageing, on the other hand of many failures by short-circuit were observed. These failures always have a local origin, and are due to defects of the components. The breakdown voltage strongly depends on the quality of the thyristor as well as the technology employed. (author)

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

Control rod drive of nuclear reactor

International Nuclear Information System (INIS)

Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

1980-01-01

This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Feedwater device for nuclear power plant

International Nuclear Information System (INIS)

Ikekita, Iwao.

1980-01-01

Purpose: To conduct water feeding without using high pressure steam of the reactor and with no radiation exposure by the provision of each feedwater pump driven by each motor controlled from variable frequency thyristor-inverter to a feedwater pipe connecting a condensate pump and the reactor. Constitution: High pressure steams resulted from heat exchange in the reactor core are transferred by way of a main steam check valve in a main steam pipe to a high pressure turbine, drive the high pressure turbine, flow out of the turbine and then drive a low pressure turbine by way of a moisture separator. The steams thus used for the turbine driving are condensed in a condensator and then sent under pressure by way of each condensating pump to a feedwater pipe. Since each of the feedwater pumps provided in the route of the feedwater pipe is driven by each of the motors under the control of the variable frequency thyristor-inverter in starting, shut down and normal operation, water is fed to the reactor. (Horiuchi, T.)

Nuclear reactor power control device

International Nuclear Information System (INIS)

Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

1987-01-01

Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

Coordinate control of integral reactor based on single neuron PID controller

International Nuclear Information System (INIS)

Liu Yan; Xia Hong

2014-01-01

As one of the main type of reactors in the future, the development of the integral reactor has attracted worldwide attention. On the basis of understanding the background of the integral reactor, the author will be familiar with and master the power control of reactor and the feedwater flow control of steam generator, and the speed control of turbine (turbine speed control is associated with the turbine load control). According to the expectative program 'reactor power following turbine load' of the reactor, it will make coordinate control of the three and come to a overall control scheme. The author will use the supervisory learning algorithm of Hebb for single neuron PID controller with self-adaptation to study the coordinate control of integral reactor. Compared with conventional PI or PID controller, to a certain extent, it solves the problems that traditional PID controller is not easy to tune real-time parameters and lack of effective control for a number of complex processes and slow-varying parameter systems. It improves the security, reliability, stability and flexibility of control process and achieves effective control of the system. (authors)

Reactor instrumentation and control

International Nuclear Information System (INIS)

Wach, D.; Beraha, D.

1980-01-01

The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

Propose Reactor Control and Monitoring System for RTP

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

2011-01-01

Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

Controlling a nuclear reactor with dropped control rods

International Nuclear Information System (INIS)

Mc Atee, K.R.; Alsop, B.H.

1987-01-01

A control system is described for a nuclear power plant including a reactor with a core having an upper portion and a lower portion and control rods which are inserted into and withdrawn from the core of the reactor vertically to control reactivity in the core. The system comprises: means to measure neutron flux separately in the upper portion and the lower portion of the reactor and to generate from such measurements a signal representative of axial distribution of power between the upper and lower portions of the reactor core; means to detect a dropped control rod in the reactor and to generate a dropped rod signal in response thereto; means to generate an axial power distribution limit signal representative of a critical axial power distribution for a dropped rod condition; means to compare the axial power distribution signal to the axial power distribution limit signal and to generate an axial power distribution out of limits signal when the axial power distribution signal exceeds the axial power distribution limit signal; and means responsive only to the presence of both the dropped rod signal and the axial power distribution out of limits signal to generate a signal for shutting the reactor down

Study and analysis of failure modes of the electrolytic capacitors and thyristors, applied to the protection system of the LHC (Large Hadron Collider); Etude et analyse des modes de defaillances des condensateurs electrolytiques a l'aluminium et des thyristors: appliquees au systeme de protection du LHC (Large Hadron Collider)

Energy Technology Data Exchange (ETDEWEB)

Perisse, F

2003-07-15

The study presented in this thesis is a contribution about the analysis of failures modes of electrolytic capacitors and thyristors. The studied components are main elements of the protection system of the superconductive magnets of the LHC. The study of the ageing of the electrolytic capacitors has shown that their reliability is strongly related to their technological characteristic. Evolution of their principal indicator of ageing (ESR) can be modeled according to different laws chosen according to their running mode. It appears that the prediction of failure of these components other than that due to wear can be only statistical taking into account the many causes of failure involving various modes of failure. In order to be able to evaluate influence of the ageing of the electrolytic capacitors on a system, simple models taking into account this parameters as well as the effective temperature of the component are proposed. An acceptable precision taking into account the simplicity of the models is obtained. The study of the thyristors has shown that these components have little drift of parameters in static ageing, on the other hand of many failures by short-circuit were observed. These failures always have a local origin, and are due to defects of the components. The breakdown voltage strongly depends on the quality of the thyristor as well as the technology employed. (author)

Reactor water level control device

International Nuclear Information System (INIS)

Hiramatsu, Yohei.

1980-01-01

Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

Study and analysis of failure modes of the electrolytic capacitors and thyristors, applied to the protection system of the LHC (Large Hadron Collider)

International Nuclear Information System (INIS)

Perisse, F.

2003-07-01

The study presented in this thesis is a contribution about the analysis of failures modes of electrolytic capacitors and thyristors. The studied components are main elements of the protection system of the superconductive magnets of the LHC. The study of the ageing of the electrolytic capacitors has shown that their reliability is strongly related to their technological characteristic. Evolution of their principal indicator of ageing (ESR) can be modeled according to different laws chosen according to their running mode. It appears that the prediction of failure of these components other than that due to wear can be only statistical taking into account the many causes of failure involving various modes of failure. In order to be able to evaluate influence of the ageing of the electrolytic capacitors on a system, simple models taking into account this parameters as well as the effective temperature of the component are proposed. An acceptable precision taking into account the simplicity of the models is obtained. The study of the thyristors has shown that these components have little drift of parameters in static ageing, on the other hand of many failures by short-circuit were observed. These failures always have a local origin, and are due to defects of the components. The breakdown voltage strongly depends on the quality of the thyristor as well as the technology employed. (author)

Power control device for nuclear reactors

International Nuclear Information System (INIS)

Kagawa, Tatsuo

1984-01-01

Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Semiconductor-machine system for controlling excitation of synchronous medium power generators

Energy Technology Data Exchange (ETDEWEB)

Vrtikapa, G

1982-01-01

A system for controlling excitation (ARP-29/1) is described which was developed at the ''Nikola Tesla'' institute (Czechoslavakia) for rebuilding the Zvornik hydroelectric plant with 30 MV X A units. The system corresponds to the modern level of automation and considers positive characteristics of existing equipment, it is easily included in a technological process, has small dimensions and is easily installed during overhaul of a electric generating plant, and it allows one to obtain good economic results. Two years of use have confirmed the high reliability and quality of the excitation. The excitation control system consists of synchronous motor, excitation system, automatic control of voltage, manual control of excitation unit, unit for automatic following and switching, relay automatic device with protection and warning. The excitation system of the generator has: thyristor rectifier, thyristor converter, a bridge with thyristor control unit, machine excitation generator, switch for demagnetization. The excitation system is supplied from an electric power network or from a three phase generator with permanent magnets.

Time-optimal control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.

1987-01-01

Control laws that permit adjustments in reactor power to be made in minimum time and without overshoot have been formulated and demonstrated. These control laws which are derived from the standard and alternate dynamic period equations, are closed-form expressions of general applicability. These laws were deduced by noting that if a system is subject to one or more operating constraints, then the time-optimal response is to move the system along these constraints. Given that nuclear reactors are subject to limitations on the allowed reactor period, a time-optimal control law would step the period from infinity to the minimum allowed value, hold the period at that value for the duration of the transient, and then step the period back to infinity. The change in reactor would therefore be accomplished in minimum time. The resulting control laws are superior to other forms of time-optimal control because they are general-purpose, closed-form expressions that are both mathematically tractable and readily implanted. Moreover, these laws include provisions for the use of feedback. The results of simulation studies and actual experiments on the 5 MWt MIT Research Reactor in which these time-optimal control laws were used successfully to adjust the reactor power are presented

Feedwater control system in BWR type reactor

International Nuclear Information System (INIS)

Tanji, Jun-ichi; Oomori, Takashi.

1980-01-01

Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

Power quality improvement in highly varying loads using thyristor-switched capacitor

Energy Technology Data Exchange (ETDEWEB)

Poshtan, M. [Petroleum Inst., Abu Dhabi (United Arab Emirates). Dept. of Electrical Engineering; Mokhtari, H.; Esmaeili, A. [Sharif Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Electrical Engineering

2007-07-01

Ordinary contactor-based-capacitor (CBC) banks may not be able to response quickly enough in highly varying electrical loads such as welding machines or arc furnace loads. Thyristor-switched capacitor (TSC) banks are therefore used to compensate for reactive power of highly varying loads. In this paper, the performance of a TSC was compared to CBC banks. The 2 systems, were also compared in terms of energy saving in transmission systems. Simulations carried out using PSCAD/EMTDC software showed that there was a considerable difference in the performance of the 2 systems. The shortcomings of existing CBC systems include slow response of mechanical switching systems; problem of switching more than one bank into the system; and, voltage/current transients during on-off switching. 3 refs., 6 tabs., 14 figs.

MAPLE-X10 reactor digital control system

International Nuclear Information System (INIS)

Deverno, M.T.; Hinds, H.W.

1991-10-01

The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

Computerized reactor power regulation with logarithmic controller

International Nuclear Information System (INIS)

Gossanyi, A.; Vegh, E.

1982-11-01

A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

Monitoring and control of anaerobic reactors

DEFF Research Database (Denmark)

Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

2003-01-01

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process mea...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

Monitoring and control of anaerobic reactors

DEFF Research Database (Denmark)

Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

2003-01-01

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

Realization of a neural algorithm by means of front-propagation in a thyristor-based hybrid system

CERN Document Server

Niedernostheide, F J; Freyd, O; Bode, M; Gorbatyuk, A V

2003-01-01

Propagating fronts are generic structures in a bistable diffusion-driven system and can be used to realize neural algorithms, as e.g., the Kohonen or the neural-gas algorithm. We present an analog-digital hybrid system based on a thyristor-like structure with several gate terminals. This structure represents the continuous part in which a propagating front, separating a region of high current density from a region of low current density, is used to control the learning process of the neural algorithm. With a system containing five neurons and five gates in a quasi one-dimensional arrangement it is demonstrated that an efficient parallel operating learning process can be realized by using the winner-take-all principle and the front propagation, i.e. exploiting the intrinsic dynamics of the semiconductor device. Finally, numerical and analytical investigations of the dependency of the front velocity and its width on the load current have been performed since these are essential parameters for improving the netw...

Realization of a neural algorithm by means of front-propagation in a thyristor-based hybrid system

International Nuclear Information System (INIS)

Niedernostheide, F.-J.; Schulze, H.-J.; Freyd, O.; Bode, M.; Gorbatyuk, A.V.

2003-01-01

Propagating fronts are generic structures in a bistable diffusion-driven system and can be used to realize neural algorithms, as e.g., the Kohonen or the neural-gas algorithm. We present an analog-digital hybrid system based on a thyristor-like structure with several gate terminals. This structure represents the continuous part in which a propagating front, separating a region of high current density from a region of low current density, is used to control the learning process of the neural algorithm. With a system containing five neurons and five gates in a quasi one-dimensional arrangement it is demonstrated that an efficient parallel operating learning process can be realized by using the winner-take-all principle and the front propagation, i.e. exploiting the intrinsic dynamics of the semiconductor device. Finally, numerical and analytical investigations of the dependency of the front velocity and its width on the load current have been performed since these are essential parameters for improving the network performance

Control rod for FBR type reactor

International Nuclear Information System (INIS)

Nakai, Koichi.

1993-01-01

In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

ADVANCED CONTROL FOR A ETHYLENE REACTOR

Directory of Open Access Journals (Sweden)

Dumitru POPESCU

2017-06-01

Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

Design of a nuclear reactor cooperative controller

International Nuclear Information System (INIS)

Alang-Rashid, N.K.; Heger, A.S.

1991-01-01

This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

Design strategy for control of inherently safe reactors

International Nuclear Information System (INIS)

Chisholm, G.H.

1984-01-01

Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

DEFF Research Database (Denmark)

Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

2012-01-01

With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...... conditions. Then, the screening of controlled variables and pairing is carried out by an assessment of the effect of the disturbances based on the closed loop disturbance gain plots. Two controlled structures are obtained and benchmarked by their capacity to reject the disturbances before the Anammox reactor....

Complete automation of nuclear reactors control

International Nuclear Information System (INIS)

Weill, J.

1955-01-01

The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

Heating control system for nuclear reactor

International Nuclear Information System (INIS)

Shinohara, Kaoru.

1981-01-01

Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

Design study of electrical power supply system for tokamak fusion power reactor

International Nuclear Information System (INIS)

1977-01-01

Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

Technique of nuclear reactors controls

International Nuclear Information System (INIS)

Weill, J.

1953-12-01

This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

A computer control system for a research reactor

International Nuclear Information System (INIS)

Crawford, K.C.; Sandquist, G.M.

1987-01-01

Most reactor applications until now, have not required computer control of core output. Commercial reactors are generally operated at a constant power output to provide baseline power. However, if commercial reactor cores are to become load following over a wide range, then centralized digital computer control is required to make the entire facility respond as a single unit to continual changes in power demand. Navy and research reactors are much smaller and simpler and are operated at constant power levels as required, without concern for the number of operators required to operate the facility. For navy reactors, centralized digital computer control may provide space savings and reduced personnel requirements. Computer control offers research reactors versatility to efficiently change a system to develop new ideas. The operation of any reactor facility would be enhanced by a controller that does not panic and is continually monitoring all facility parameters. Eventually very sophisticated computer control systems may be developed which will sense operational problems, diagnose the problem, and depending on the severity of the problem, immediately activate safety systems or consult with operators before taking action

ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

Directory of Open Access Journals (Sweden)

Vyacheslav K. Mayevski

2014-01-01

Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

MODELLING AND CONTROL OF CONTINUOUS STIRRED TANK REACTOR WITH PID CONTROLLER

Directory of Open Access Journals (Sweden)

Artur Wodołażski

2016-09-01

Full Text Available This paper presents a model of dynamics control for continuous stirred tank reactor (CSTR in methanol synthesis in a three-phase system. The reactor simulation was carried out for steady and transient state. Efficiency ratio to achieve maximum performance of the product per reactor unit volume was calculated. Reactor dynamics simulation in closed loop allowed to received data for tuning PID controller (proportional-integral-derivative. The results of the regulation process allow to receive data for optimum reactor production capacity, along with local hot spots eliminations or temperature runaway.

Control rod for HTGR type reactor

International Nuclear Information System (INIS)

Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

1990-01-01

Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

Method and device for controlling reactor power

International Nuclear Information System (INIS)

Oohashi, Masahisa; Masuda, Hiroyuki.

1982-01-01

Purpose: To enable load following-up operation of a reactor adapted to perform power conditioning by the control of the liquid poison density in the core and by the control rods. Constitution: In a case where the reactor power is repeatedly changed in a reactor having a liquid poison density control device and control rods, the time period for the power control is divided depending on the magnitude of the change with time in the reactivity and the optimum values are set for the injection and removal amount of the liquid poison within the divided period. Then, most parts of the control required for the power change are alloted to the liquid poison that gives no effect on the power distribution while minimizing the movement of the control rods, whereby the power change in the reactor as in the case of the load following-up operation can be practiced with ease. (Kawakami, Y.)

Fixed-bed Reactor Dynamics and Control - A Review

DEFF Research Database (Denmark)

Jørgensen, S. B.

1986-01-01

The industrial diversity of fixed bed reactors offers a challenging and relevant set of control problems. These intricate problems arise due to the rather complex dynamics of fixed bed reactors and to the complexity of actual reactor configurations. Many of these control problems are nonlinear...... and multi-variable. During the last decade fixed bed reactor control strategies have been proposed and investigated experimentally. This paper reviews research on these complex control problems with an emphasis upon solutions which have been demon-strated to work in the laboratory and hold promise...

Nuclear reactor power control system based on flexibility model

International Nuclear Information System (INIS)

Li Gang; Zhao Fuyu; Li Chong; Tai Yun

2011-01-01

Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective. (author)

Reactor feedwater control device

International Nuclear Information System (INIS)

Koshi, Yuji.

1993-01-01

In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

Design of controller for control rod of research reactors

International Nuclear Information System (INIS)

Abou-Zaid, R.M.F.M

2008-01-01

Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this thesis, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In-The-Loop (HIL). Some of the control loop components are real hardware components and the others are simulated. First, the whole system is modeled and tested by Real-Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-loop simulation is tested using Real-Time Windows Target in MATLAB and Visual C ++ . The control parts are included as hardware components which are the reactor control rod and its drivers. Three kinds of controllers are studied, Proportional-Derivative (PD), Proportional-Integral-Derivative (PID) and Fuzzy controller. An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain.

The resonance absorption controlled reactor

Energy Technology Data Exchange (ETDEWEB)

Caro, R

1977-07-01

In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D{sub 2}O/H{sub 2}O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs.

The resonance absorption controlled reactor

International Nuclear Information System (INIS)

Caro, R.

1977-01-01

In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D 2 O/H 2 O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs

Design of Multi Objectives Control Systems to Control Nuclear Reactor Power

International Nuclear Information System (INIS)

Abdelaal, M.M.Z.

2013-01-01

The Egyptian Testing Research Reactor (ETRR-2) nonlinear twelfth order model is linearized and reduced to lower order model. Model order reduction methodologies such as balanced truncation, Schur reduction method, Hankel approximation and Coprime factorization have been used in the reduction process. The reactor actually controlled by PD controller with fixed tuning parameters. LMI state feedback, LMI-pool assignment, H ∞ and observer based controllers based third order model are proposed to be used in the reactor power control instead of the PD controller. A comparison of LMI, LMI-Pole placement,H ∞ control systems and those of based observer relative to the PD controller has been performed which showed better response and disturbance rejection for the proposed controllers.

KS-150 reactor control

International Nuclear Information System (INIS)

Wagner, K.

1974-01-01

A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

Control Rod Malfunction at the NRAD Reactor

Energy Technology Data Exchange (ETDEWEB)

Thomas L. Maddock

2010-05-01

The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

Fission reactor control rod

International Nuclear Information System (INIS)

Irie, Tomoo.

1991-01-01

The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

1993-01-01

Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

DEFF Research Database (Denmark)

Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

2012-01-01

With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...

Dynamics and Control of Chemical Reactors-Selectively Surveyed

DEFF Research Database (Denmark)

Jørgensen, S. B.; Jensen, N.

1989-01-01

The chemical reactor or bioreactor is physically at a central position in a process, and often with a decisive role on the overall technical and economical performance. Even though application of feedback control on reactors is gaining momentum and on-line optimization has been implemented....... For bioreactors the theory and practice of reactor design, dynamics and control have to be adapted to the peculiarities of the biological catalysts. Enzymes, the protein catalysts, are the simplest ones, which have many common features with chemical catalysts. The living cells are much more complex, these growing...... in industry, many reactor control problems are still left unsolved or only partly solved using open loop strategies where disturbance rejection and model inaccuracies have to be handled through manual reactor control and feedback control of raw material preprocessing and product purification operations...

Adaptive robust control of the EBR-II reactor

International Nuclear Information System (INIS)

Power, M.A.; Edwards, R.M.

1996-01-01

Simulation results are presented for an adaptive H ∞ controller, a fixed H ∞ controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H ∞ controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H ∞ and classical controllers. This makes for a superior and more robust controller

Distributed expert systems for nuclear reactor control

International Nuclear Information System (INIS)

Otaduy, P.J.

1992-01-01

A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

The design and construction of a controllable reactor with a HTS control winding

International Nuclear Information System (INIS)

Wass, Torbjoern; Hoernfeldt, Sven; Valdemarsson, Stefan

2006-01-01

Reactive power compensation is vital for obtaining efficient operation of long transmission power lines or cables. The need of reactive power changes with the load of the transmission line. Discrete units of conventional reactors are therefore switched in and out in order to obtain more efficient reactive power compensation. A continuous reactive compensation will reduce the transmission losses and increase the transmission capacity of active power. We have designed and constructed a one phase small scale prototype of a controllable shunt reactor with a high temperature superconducting control winding. The reactor consists basically of two windings and an iron core. The control winding is placed so that it generates a DC magnetic field perpendicular to the main AC magnetic field. Thus the DC current in the control winding can control the direction of the magnetization of the iron core and thereby the reactance of the reactor. Such a control winding will have low losses and give the reactor a large dynamic range. For this small scale reactor we found that the reactive power could be varied with a factor six. We have demonstrated the feasibility to design large scale controllable shunt reactors with large dynamic range and low losses utilizing a control winding made of a high temperature superconductor

Application of fuzzy logic control system for reactor feed-water control

International Nuclear Information System (INIS)

Iijima, T.; Nakajima, Y.

1994-01-01

The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

Micro processor based research reactor instrumentation and control system

International Nuclear Information System (INIS)

Hyde, W.K.

1987-01-01

The system consists of a Control System Computer (CSC) incorporated into a Reactor Control Console (RCC) and a Data Acquisition and Control Unit (DAC) adjacent to the reactor. The CSC has a high resolution color graphics CRT monitor which provides real-time graphic simulation of the reactor and a number of bar graphs displaying strategic parameters of the reactor system. In addition, abnormal or dangerous conditions are displayed. The CSC is equipped with two printers eliminating manual logging of reactor data. The reactor display and pulse mode display may also be printed. Historical data is saved in the system's large capacity memory and may be replayed and/or printed. Because of the CSC's inherent high speed math capability, raw reactor data will be quickly converted and displayed in real-time. Data can be presented in meaningful engineering units. The DAC provides a high speed data acquisition and control capability adjacent to the reactor. It continuously collects data from the reactor system, concentrates the data into a database and transmits it to the CSC when requested. Data transmission is over one of two data trunks to the CSC. The secondary trunk is used if the primary trunk fails. The data trunks drastically reduce the wiring requirements between the reactor and the Control Console. During steady-state operation of the reactor, operator commands to adjust the rod positions is transmitted from the CSC to the DAC which re-issues the commands to the drive mechanisms. In the automatic mode, the DAC will control the position of the rods via a PID algorithm. The system is independently monitored by two or more safety computers. Their function is to monitor the power level, the rate of change of power and fuel temperature of the reactor and to independently shut the reactor down in the event of a potentially dangerous (scram) condition. (author)

Adaptive control method for core power control in TRIGA Mark II reactor

Science.gov (United States)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Method of controlling the reactor operation

International Nuclear Information System (INIS)

Ishiguro, Akira; Nakakura, Hiroyuki.

1987-01-01

Purpose: To moderate vibratory response due to delayed operation thereby obtain stable controlled response in the operation control for a PWR type reactor. Method: the reactor operation is controlled by the axial power distribution control by regulating the boron concentration in primary coolants with a boron density control system and controlling the average temperature for the primary coolants with the control rod control system. In this case, the control operation and the control response become instable due to transmission delay, etc. of aqueous boric acid injection to the primary coolant circuits to result in vibratory response. In the present invention, signals are prepared by adding the amount in proportion to the variation coefficient with time of xenone concentration obtained from the measured value for the reactor power added to the conventional axial power distribution parameter deviation and used as the input signals for the boron concentration control system. As a result, the instability due to the transmission delay of the aqueous boric acid injection is improved by the preceding control by the amount in proportion with the variation coefficient with time of the xenone concentration. An advantageous effect can be expected for the load following operation during day time according to the present invention. (Kamimura, M.)

Computer simulation of nuclear reactor control by means of heuristic learning controller

International Nuclear Information System (INIS)

Bubak, M.; Moscinski, J.

1976-01-01

A trial of application of two techniques of Artificial Intelligence: heuristic Programming and Learning Machines Theory for nuclear reactor control is presented. Considering complexity of the mathematical models describing satisfactorily the nuclear reactors, value changes of these models parameters in course of operation, knowledge of some parameters value with too small exactness, there appear diffucluties in the classical approach application for these objects control systems design. The classical approach consists in definition of the permissible control actions set on the base of the set performance index and the object mathematical model. The Artificial Intelligence methods enable construction of the control system, which gets during work an information being a priori inaccessible and uses it for its action change for the control to be the optimum one. Applying these methods we have elaborated the reactor power control system. As the performance index there has been taken the integral of the error square. For the control system there are only accessible: the set power trajectory, the reactor power and the control rod position. The set power trajectory has been divided into time intervals called heuristic intervals. At the beginning of every heuristic interval, on the base of the obtained experience, the control system chooses from the control (heuristic) set the optimum control. The heuristic set it is the set of relations between the control rod rate and the state variables, the set and the obtained power, similar to simplifications applied by nuclear reactors operators. The results obtained for the different control rod rates and different reactor (simulated on the digital computer) show the proper work of the system. (author)

Centralized digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, K.C.

1987-01-01

A hardware and software design for the centralized control of a research nuclear reactor by a digital computer are presented, as well as an investigation of automatic-feedback control. Current reactor-control philosophies including redundancy, inherent safety in failure, and conservative-yet-operational scram initiation were used as the bases of the design. The control philosophies were applied to the power-monitoring system, the fuel-temperature monitoring system, the area-radiation monitoring system, and the overall system interaction. Unlike the single-function analog computers currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control-rod movements to conform with operator requests, automatically log the required physical parameters during reactor operation, perform the required system tests, and monitor facility safety and security. Reactor power control is based on signals received from ion chambers located near the reactor core. Absorber-rod movements are made to control the rate of power increase or decrease during power changes and to control the power level during steady-state operation. Additionally, the system incorporates a rudimentary level of artificial intelligence

Neutron transmutation doping technology of silicon and overview of trial irradiations at Cirus reactor

International Nuclear Information System (INIS)

Singh, Tej; Bhatnagar, Anil; Singh, Kanchhi; Raina, V.K.

2007-12-01

Neutron transmutation doped silicon (NTD-Si) has been used extensively in manufacturing of high power semiconductor devices. The quality of NTD-Si, both from view points of dopant concentration and homogeneity has been found superior to the quality of doped silicon produced by conventional methods. The technology of NTD-Si has been perfected to achieve more accurate resistivity and homogenous resistivity with complete elimination of hot spots. In addition, the greater spatial uniformity, as well as the precise control over the resistivity achievable by using the NTD process, has led to a substantial increase in the breakdown voltage capability of thyristors. The report describes the fundamentals of NTD-Si production and discusses various techniques used for control of dopant concentration and homogeneity. Various aspects like radiation damage, residual radio-activity, nuclear heating, surface contamination and annealing requirements of the silicon ingots after irradiation have also been discussed. Details of trail irradiation and characterization of NTD-Si samples have been provided. Future plans for production of NTD-Si in Cirus and Dhruva reactors have also been discussed. (author)

REACTOR CONTROL ROD OPERATING SYSTEM

Science.gov (United States)

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

Digital control for nuclear reactors - lessons learned

International Nuclear Information System (INIS)

Bernard, J.A.; Aviles, B.N.; Lanning, D.D.

1992-01-01

Lessons learned during the course of the now decade-old MIT program on the digital control of nuclear reactors are enumerated. Relative to controller structure, these include the importance of a separate safety system, the need for signal validation, the role of supervisory algorithms, the significance of command validation, and the relevance of automated reasoning. Relative to controller implementation, these include the value of nodal methods to the creation of real-time reactor physics and thermal hydraulic models, the advantages to be gained from the use of real-time system models, and the importance of a multi-tiered structure to the simultaneous achievement of supervisory, global, and local control. Block diagrams are presented of proposed controllers and selected experimental and simulation-study results are shown. In addition, a history is given of the MIT program on reactor digital control

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Application of H∞ control theory to power control of a nonlinear reactor model

International Nuclear Information System (INIS)

Suzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni

1993-01-01

The H∞ control theory is applied to the compensator design of a nonlinear nuclear reactor model, and the results are compared with standard linear quadratic Gaussian (LQG) control. The reactor model is assumed to be provided with a control rod drive system having the compensation of rod position feedback. The nonlinearity of the reactor model exerts a great influence on the stability of the control system, and hence, it is desirable for a power control system of a nuclear reactor to achieve robust stability and to improve the sensitivity of the feedback control system. A computer simulation based on a power control system synthesized by LQG control was performed revealing that the control system has some stationary offset and less stability. Therefore, here, attention is given to the development of a methodology for robust control that can withstand exogenous disturbances and nonlinearity in view of system parameter changes. The developed methodology adopts H∞ control theory in the feedback system and shows interesting features of robustness. The results of the computer simulation indicate that the feedback control system constructed by the developed H∞ compensator possesses sufficient robustness of control on the stability and disturbance attenuation, which are essential for the safe operation of a nuclear reactor

Level controlling system in BWR type reactors

International Nuclear Information System (INIS)

Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

1981-01-01

Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

Regulatory Framework for Controlling the Research Reactor Decommissioning Project

International Nuclear Information System (INIS)

Melani, Ai; Chang, Soon Heung

2009-01-01

Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

An autonomous control framework for advanced reactors

Directory of Open Access Journals (Sweden)

Richard T. Wood

2017-08-01

Full Text Available Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

An autonomous control framework for advanced reactors

International Nuclear Information System (INIS)

Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C.

2017-01-01

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors

An autonomous control framework for advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

Reactor power control method and device

International Nuclear Information System (INIS)

Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

1997-01-01

The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

The reactor power control system based on digital control in nuclear power plant

International Nuclear Information System (INIS)

Liu Chong; Zhou Jianliang; Tan Ping

2010-01-01

The PLC (Programmable Logical Controller), digital communication and redundant techniques are applied in the rod control and position indication system(namely the reactor power control system) to perform the power control in the 300 MW reactor automatically and integrally in Qinshan Phase I project. This paper introduces the features, digital design methods of hardware of the instrumentation and control system (I and C) in the reactor power control. It is more convenient for the information exchange by human-machine interface (HMI), operation and maintenance, and the system reliability has been greatly improved after the project being reconstructed. (authors)

Automatic power control for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Yung Joon

1994-02-01

During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

Utilization of the research reactors for the power reactor control instrumentation development

International Nuclear Information System (INIS)

Duchene, J.; Verdant, R.; Gilbert, J.

1977-01-01

Studies on characteristics and reliability of control instruments lead to testing with various radiations of various intensities and energy spectra. Osiris and Triton reactors present this great variety of radiations and a flexibility of use better than power reactors [fr

Control aid for xenon vibration in reactor

International Nuclear Information System (INIS)

Kanekawa, Takashi.

1990-01-01

In the present invention, the control operation for suppressing xenon vibrations in a reactor is aided for saving forecasting analysis and operator's skills. That is, parameters to be controlled for the suppression of xenon vibrations are power distribution, iodine distribution and xenon distribution. But what can be observed by operaters by the conventional fast overtone method is only the output distribution. In the present invention, the output level of the reactor core is always observed. Then, mathematical processings are conducted for the iodine distribution, the xenon distribution and the power distribution in the reactor core based on the histeresis of the parameters obtained by the measurement using physical constants and reactor design data. The xenon vibration control is aided by displaying the change with time of the distortion in axial direction. Accordingly, operators can always recognize the axial distortion of the power distribution, the iodine distribution and the xenon distribution. (I.S.)

Control rod for the operation of nuclear reactor

International Nuclear Information System (INIS)

Ishida, Hiromi

1987-01-01

Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

Method of changing the control rod pattern in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to change the control rod pattern in a short time with ease, as well as improve the availability factor of the reactor. Method: Control rods other than those being inserted into the reactor core are inserted into the reactor core to reduce the power by the reduction in the reactor core flow rate. Then, the control rod to be operated is operated partially for the change of the control rod pattern to restrict the linear heat rating of the fuels to less than 0.1 kW/ft per one hour to change the control pattern to the aimed control rod pattern. Then, the reactor core flow rate is increased after the pattern exchange for the control rod to increase the power. Since only the control rod operation is performed without adjusting the reactor core flow rate upon change of the control rod pattern, procedures can be made simply in a short time to thereby improve the availability factor of the reactor. (Moriyama, K.)

Comparing Effects of TCR and TSC on MHO Distance Protection Setting in 400 kV Algerian Transmission Line

Directory of Open Access Journals (Sweden)

Mohamed ZELLAGUI

2012-11-01

Full Text Available This paper presents a study on the performances of distance relays setting in 400 kV in Eastern Algerian transmission networks at Sonelgaz Group (Algerian company of Electrical and Gas compensated by shunt Flexible AC Transmission System (FACTS. The facts are used for controlling transmission voltage, power flow, reactive power, and damping of power system oscillations in high power transfer levels. The effects of SVC devices i.e. Thyristor Controlled Reactor (TCR and the Thyristor Switched Capacitors (TSC insertion, on the total impedance of a transmission line protected by MHO distance relay are investigated. The modified setting zones protections for three forward zones (Z1, Z2 and Z3 have been calculated in order to improve the performances of distance relay protection and prevent circuit breaker nuisance tripping.

Power controlling method for BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1983-01-01

Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

International Nuclear Information System (INIS)

Ware, A.G.

1988-01-01

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

Directory of Open Access Journals (Sweden)

Alexander Labutin

2017-02-01

Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

Science.gov (United States)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-11-01

High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Development of Power Controller System based on Model Reference Adaptive Control for a Nuclear Reactor

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Ridzuan Abdul Mutalib

2014-01-01

The Reactor TRIGA PUSPATI (RTP)-type TRIGA Mark II was installed in the year 1982. The Power Controller System (PCS) or Automated Power Controller System (APCS) is very important for reactor operation and safety reasons. It is a function of controlled reactivity and reactor power. The existing power controller system is under development and due to slow response, low accuracy and low stability on reactor power control affecting the reactor safety. The nuclear reactor is a nonlinear system in nature, and it is power increases continuously with time. The reactor parameters vary as a function of power, fuel burnup and control rod worth. The output power value given by the power control system is not exactly as real value of reactor power. Therefore, controller system design is very important, an adaptive controller seems to be inevitable. The method chooses is a linear controller by using feedback linearization, for example Model Reference Adaptive Control. The developed APCS for RTP will be design by using Model Reference Adaptive Control (MRAC). The structured of RTP model to produce the dynamic behaviour of RTP on entire operating power range from 0 to 1MWatt. The dynamic behavior of RTP model is produced by coupling of neutronic and thermal-hydraulics. It will be developed by using software MATLAB/Simulink and hardware module card to handle analog input signal. A new algorithm for APCS is developed to control the movement of control rods with uniformity and orderly for RTP. Before APCS test to real plant, simulation results shall be obtained from RTP model on reactor power, reactivity, period, control rod positions, fuel and coolant temperatures. Those data are comparable with the real data for validation. After completing the RTP model, APCS will be tested to real plant on power control system performance by using real signal from RTP including fail-safe operation, system reliable, fast response, stability and accuracy. The new algorithm shall be a satisfied

Method of driving control rod in reactor

International Nuclear Information System (INIS)

Osa, Hirotaka.

1986-01-01

Purpose: To improve security and safety of the reactor by reducing reactor output automatically and quickly when circulation of cooling water is stopped. Constitution: When the circulating pump is under operation, fluid pressure in the discharge pipe is transferred to the fluid room of fluid pressure cylinder via the control rod drive pipe and lift up the piston, and then the control rod is drawn out of the reactor core. When the circulating pump is lowered in its functions, discharge pipe fluid pressure decreases, fluid pressure in the fluid room decreases, and with less force of piston movement, the control rod gets lowered by its own weight. At this time, the blocked state of the opening by the piston is released, fluid flows into the room. Lowering of pressure and the control rod is promoted by transferring out fluid below the piston in the fluid room to the upper part of the piston via a small gap when the control rod falls by gravity. (Horiuchi, T.)

Development of the 'JFT-2' tokamak plasma position control system

International Nuclear Information System (INIS)

Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

1980-01-01

Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

Nuclear reactor shutdown control rod assembly

International Nuclear Information System (INIS)

Bilibin, K.

1988-01-01

This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

a microprocessor-based control, scheme for a pwm voltage-fed

African Journals Online (AJOL)

Dr Obe

power circuit over-current and over-voltage protection against adverse changes in the inverter input supply and the ... integral controllers constitute the main control program while the application of thyristor gating ... current flow to the load.

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

Power control device in nuclear reactor

International Nuclear Information System (INIS)

Koyama, Kazuaki.

1981-01-01

Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

Control rod drives for FBR type reactor

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1990-01-01

The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

Protective guide structure for reactor control rod

International Nuclear Information System (INIS)

Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

1996-01-01

The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

Optimization and control of a continuous polymerization reactor

Directory of Open Access Journals (Sweden)

L. A. Alvarez

2012-12-01

Full Text Available This work studies the optimization and control of a styrene polymerization reactor. The proposed strategy deals with the case where, because of market conditions and equipment deterioration, the optimal operating point of the continuous reactor is modified significantly along the operation time and the control system has to search for this optimum point, besides keeping the reactor system stable at any possible point. The approach considered here consists of three layers: the Real Time Optimization (RTO, the Model Predictive Control (MPC and a Target Calculation (TC that coordinates the communication between the two other layers and guarantees the stability of the whole structure. The proposed algorithm is simulated with the phenomenological model of a styrene polymerization reactor, which has been widely used as a benchmark for process control. The complete optimization structure for the styrene process including disturbances rejection is developed. The simulation results show the robustness of the proposed strategy and the capability to deal with disturbances while the economic objective is optimized.

Device for rearranging control rods of experimental reactors

International Nuclear Information System (INIS)

Louda, J.

1975-01-01

The invention claims a means for the adjustment of control rods in experimental reactors with a continuously variable pitch of the fuel element spacer. The proposed device permits obtaining maximum variability in the physical modelling of nuclear power reactor cores in experimental reactors. (F.M.)

Nuclear reactor kinetics and plant control

CERN Document Server

Oka, Yoshiaki

2013-01-01

Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

Liquid-poison type power controlling device for nuclear reactor

International Nuclear Information System (INIS)

Horiuchi, Tetsuo; Yamanari, Shozo; Sugisaki, Toshihiko; Goto, Hiroshi.

1981-01-01

Purpose: To improve the safety and the operability of a nuclear reactor by adjusting the density of liquid poison. Constitution: The thermal expansion follow-up failure between cladding and a pellet upon abrupt and local variations of the power is avoided by adjusting the density of liquid poison during ordinary operation in combination with a high density liquid poison tank and a filter and smoothly controlling the reactor power through a pipe installed in the reactor core. The high density liquid poison is abruptly charged in to the reactor core under relatively low pressure through the tube installed in the reactor core at the time of control rod insertion failure in an accident, thereby effectively shutting down the reactor and improving the safety and the operability of the reactor. (Yoshihara, H.)

Fast switching thyristor applied in nanosecond-pulse high-voltage generator with closed transformer core.

Science.gov (United States)

Li, Lee; Bao, Chaobing; Feng, Xibo; Liu, Yunlong; Fochan, Lin

2013-02-01

For a compact and reliable nanosecond-pulse high-voltage generator (NPHVG), the specification parameter selection and potential usage of fast controllable state-solid switches have an important bearing on the optimal design. The NPHVG with closed transformer core and fast switching thyristor (FST) was studied in this paper. According to the analysis of T-type circuit, the expressions for the voltages and currents of the primary and secondary windings on the transformer core of NPHVG were deduced, and the theoretical maximum analysis was performed. For NPHVG, the rise-rate of turn-on current (di/dt) across a FST may exceed its transient rating. Both mean and maximum values of di/dt were determined by the leakage inductances of the transformer, and the difference is 1.57 times. The optimum winding ratio is helpful to getting higher voltage output with lower specification FST, especially when the primary and secondary capacitances have been established. The oscillation period analysis can be effectively used to estimate the equivalent leakage inductance. When the core saturation effect was considered, the maximum di/dt estimated from the oscillating period of the primary current is more accurate than one from the oscillating period of the secondary voltage. Although increasing the leakage inductance of NPHVG can decrease di/dt across FST, it may reduce the output peak voltage of the NPHVG.

Self-teaching neural network learns difficult reactor control problem

International Nuclear Information System (INIS)

Jouse, W.C.

1989-01-01

A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits

Instrumentation and control strategies for an integral pressurized water reactor

Directory of Open Access Journals (Sweden)

Belle R. Upadhyaya

2015-03-01

Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

The control of reactor outages

International Nuclear Information System (INIS)

Bouget, Y.H.; Berteloot, J.M.

1995-01-01

The 1985-1992 period was marked by a continuous decay in French reactors operation. This situation has led the Committee for Outages Mastery to take steps for the improvement of nuclear power plants availability. The control of reactor outages requires an integrated vision of the safety, duration, dosimetry, costs and security aspects and a perfect management of contractors. The paper describes the methodology used for the management and the maintenance of the French PWR reactors stock. A detailed schedule of maintenance tasks with dose estimations is now required from each site to anticipate and optimize the duration of outages. Thanks to this action, a significant reduction of the maintenance costs is observed for the 1992-1995 period. (J.S.). 2 figs

Status of fast reactor control rod development in the United Kingdom

International Nuclear Information System (INIS)

Kelly, B.T.

1984-01-01

The two large fast reactors constructed in the United Kingdom, that is the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR) differed substantially in their control systems. DFR was controlled by variation of the neutron leakage from the core while PFR uses conventional control rods containing neutron absorbing materials. This paper describes the development of the PFR control systems, the progressive design of the control systems for the prototype Civil Fast Reactor (CFR) and the supporting research and development programmes. (author)

Computer Controlled Chemical Micro-Reactor

International Nuclear Information System (INIS)

Mechtilde, Schaefer; Eduard, Stach; Adreas, Foitzik

2006-01-01

Chemical reactions or chemical equilibria can be influenced and controlled by several parameters. The ratio of two liquid ingredients, the so called reactants or educts, plays an important role in determining the end product and its yield. The reactants must be weighed and accordingly mixed with the conventional batch mode. If the reaction is done in a microreactor or in several parallel working micro-reactors, units for allotting the educts in appropriate quantities are required. In this report we present a novel micro-reactor that allows the constant monitoring of the chemical reaction via Raman spectroscopy. Such monitoring enables an appropriate feedback on the steering parameters for the PC controlled micro-pumps for the appropriate educt flow rate of both liquids to get optimised ratios of ingredients at an optimised total flow rate. The micro-reactors are the core pieces of the design and are easily removable and can therefore be changed at any time to adapt the requirements of the chemical reaction. One type of reactor consists of a stainless steel base containing small scale milled channels covered with anodically bonded Pyrex glass. Another type of reactor has a base of anisotropically etched silicon, and is also covered with anodically bonded Pyrex glass. The glass window allows visual observation of the initial phase interface of the two educts in the reaction channels by optical microscopy and does not affect, in contrast to infrared spectroscopy, the Raman spectroscopic signal for detection of the reaction kinetics. On the basis of a test reaction, we present non-invasive and spatially highly resolved in-situ reaction analysis using Raman spectroscopy measured along the reaction channel at different locations

The optimal control of ITU TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Can, Burhanettin

2008-01-01

In this study, optimal control of ITU TRIGA Mark-II Reactor is discussed. A new controller has been designed for ITU TRIGA Mark-II Reactor. The controller consists of main and auxiliary controllers. The form is based on Pontragyn's Maximum Principle and the latter is based on PID approach. For the desired power program, a cubic function is chosen. Integral Performance Index includes the mean square of error function and the effect of selected period on the power variation. YAVCAN2 Neutronic - Thermal -Hydraulic code is used to solve the equations, namely 11 equations, dealing with neutronic - thermal - hydraulic behavior of the reactor. For the controller design, a new code, KONTCAN, is written. In the application of the code, it is seen that the controller controls the reactor power to follow the desired power program. The overshoot value alters between 100 W and 500 W depending on the selected period. There is no undershoot. The controller rapidly increases reactivity, then decreases, after that increases it until the effect of temperature feedback is compensated. Error function varies between 0-1 kW. (author)

Control parameter optimization for AP1000 reactor using Particle Swarm Optimization

International Nuclear Information System (INIS)

Wang, Pengfei; Wan, Jiashuang; Luo, Run; Zhao, Fuyu; Wei, Xinyu

2016-01-01

Highlights: • The PSO algorithm is applied for control parameter optimization of AP1000 reactor. • Key parameters of the MSHIM control system are optimized. • Optimization results are evaluated though simulations and quantitative analysis. - Abstract: The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. The MSHIM core control system can provide superior reactor control capabilities via automatic rod control only. This enables the AP1000 to perform power change operations automatically without the soluble boron concentration adjustments. In this paper, the Particle Swarm Optimization (PSO) algorithm has been applied for the parameter optimization of the MSHIM control system to acquire better reactor control performance for AP1000. System requirements such as power control performance, control bank movement and AO control constraints are reflected in the objective function. Dynamic simulations are performed based on an AP1000 reactor simulation platform in each iteration of the optimization process to calculate the fitness values of particles in the swarm. The simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model and the MSHIM control strategy. Based on the simulation platform, the typical 10% step load decrease transient from 100% to 90% full power is simulated and the objective function used for control parameter tuning is directly incorporated in the simulation results. With successful implementation of the PSO algorithm in the control parameter optimization of AP1000 reactor, four key parameters of the MSHIM control system are optimized. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability and reduce the control rod movement without compromising AO control. Therefore, the PSO based optimization

Nuclear reactor plants and control systems therefor

International Nuclear Information System (INIS)

de Boer, G.A.; de Hex, M.

1976-01-01

A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

Digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, Kevan

1986-01-01

Currently, the use of digital computers in energy producing systems has been limited to data acquisition functions. These computers have greatly reduced human involvement in the moment to moment decision process and the crisis decision process, thereby improving the safety of the dynamic energy producing systems. However, in addition to data acquisition, control of energy producing systems also includes data comparison, decision making, and control actions. The majority of the later functions are accomplished through the use of analog computers in a distributed configuration. The lack of cooperation and hence, inefficiency in distributed control, and the extent of human interaction in critical phases of control have provided the incentive to improve the later three functions of energy systems control. Properly applied, centralized control by digital computers can increase efficiency by making the system react as a single unit and by implementing efficient power changes to match demand. Additionally, safety will be improved by further limiting human involvement to action only in the case of a failure of the centralized control system. This paper presents a hardware and software design for the centralized control of a research nuclear reactor by a digital computer. Current nuclear reactor control philosophies which include redundancy, inherent safety in failure, and conservative yet operational scram initiation were used as the bases of the design. The control philosophies were applied to the power monitoring system, the fuel temperature monitoring system, the area radiation monitoring system, and the overall system interaction. Unlike the single function analog computers that are currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control rod movements to conform with operator requests, automatically log the required physical parameters during reactor

Modelling and control design for SHARON/Anammox reactor sequence

DEFF Research Database (Denmark)

Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

2012-01-01

metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure......With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...

Neutron density optimal control of A-1 reactor analoque model

International Nuclear Information System (INIS)

Grof, V.

1975-01-01

Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

Digital control for the Penn State Breazeale reactor

International Nuclear Information System (INIS)

Raiskums, G.A.

1991-01-01

Digital control has been an integral part of Canada deuterium uranium (CANDU) nuclear power reactor technology since the 1960s. Much of the high CANDU production reliability can be attributed to the fault-tolerant and flexible control algorithms achievable with digital control. Atomic Energy of Canada Limited (AECL) has now transported this technology to research reactors, using industrial-grade microcomputers to solve equipment aging and spares obsolescence problems so prevalent at older installations. The open architecture of the Intel 8086-based computers provides for wide availability and reasonably priced, quality hardware from numerous sources. AECL recently supplied the Pennsylvania State University Breazeale Reactor (PSBR) with a new console containing a digital control and monitoring system. The reactor safety system (RSS) was also replaced with hardwired relay logic and truly analog state-of-the-art wide range nuclear instrumentation supplied by AECL's subcontractor, Gamma-Metrics. Retaining analog hardware for the mandated RSS functions was key to minimizing licensing efforts and the extensive verification and validation that would be required for safety system software. This paper elaborates on the digital control and monitoring portion of the PSBR console replacement, with emphasis on the key system objectives

Corrosion control in CANDU nuclear power reactors

International Nuclear Information System (INIS)

Lesurf, J.E.

1974-01-01

Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

MATLAB/SIMULINK model of CANDU reactor for control studies

International Nuclear Information System (INIS)

Javidnia, H.; Jiang, J.

2006-01-01

In this paper a MATLAB/SIMULINK model is developed for a CANDU type reactor. The data for the reactor are taken from an Indian PHWR, which is very similar to CANDU in its design. Among the different feedback mechanisms in the core of the reactor, only xenon has been considered which plays an important role in spatial oscillations. The model is verified under closed loop scenarios with simple PI controller. The results of the simulation show that this model can be used for controller design and simulation of the reactor systems. Adding models of the other components of a CANDU reactor would ultimately result in a complete model of CANDU plant in MATLAB/SIMULINK. (author)

MATLAB/SIMULINK platform for simulation of CANDU reactor control system

International Nuclear Information System (INIS)

Javidnia, H.; Jiang, J.

2007-01-01

In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

Dynamics and control of molten-salt breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Dynamics and control of molten-salt breeder reactor

Directory of Open Access Journals (Sweden)

Vikram Singh

2017-08-01

Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

Description of a research reactor control system using a programmable controller

International Nuclear Information System (INIS)

Battle, R.E.

1986-01-01

This paper describes the design features, testing methods, and operational experience of a programmable controller (PC) installed as a neutron flux controller in the Oak Ridge Research Reactor (ORR) at Oak Ridge National Laboratory (ORNL). The PC was designed to control neutron flux from 1 to 105% for three selectable ranges. The PC generates a flux setpoint under operator control, calculates the reactor heat power from flow and temperature signals, calculates a neutron flux calibration factor based on the heat power, and positions a control rod based on the flux-setpoint difference. The programmable controller was tested by controlling an analog computer model of the ORR. The equipment was installed in August 1985, and except for some startup problems, the system has performed well

Status of control assembly materials in Indian water reactors

International Nuclear Information System (INIS)

Date, V.G.; Kulkarni, P.G.

2000-01-01

India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

Modern control technology for improved nuclear reactor performance

International Nuclear Information System (INIS)

Oakes, L.C.

1986-01-01

One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level

Three-dimensional harmonic control of a nuclear reactor

International Nuclear Information System (INIS)

Potapenko, P.T.

1989-01-01

Algorithms for neutron flux control based on harmonic three-dimensional core are considered. The essence of the considered approach includes determination of harmonics amplitudes by signals self-powered detectors placed in reactor channels and reconstruction of neutron field distribution over the reactor core volume using the data obtained. Neutron field harmonic control is shown to be reduced to independent measurement and calculation of height harmonics in channels using techniques developed for channel power control

Feed water control device in a reactor

International Nuclear Information System (INIS)

Okutani, Tetsuro.

1984-01-01

Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

Wide-range nuclear reactor temperature control using automatically tuned fuzzy logic controller

International Nuclear Information System (INIS)

Ramaswamy, P.; Edwards, R.M.; Lee, K.Y.

1992-01-01

In this paper, a fuzzy logic controller design for optimal reactor temperature control is presented. Since fuzzy logic controllers rely on an expert's knowledge of the process, they are hard to optimize. An optimal controller is used in this paper as a reference model, and a Kalman filter is used to automatically determine the rules for the fuzzy logic controller. To demonstrate the robustness of this design, a nonlinear six-delayed-neutron-group plant is controlled using a fuzzy logic controller that utilizes estimated reactor temperatures from a one-delayed-neutron-group observer. The fuzzy logic controller displayed good stability and performance robustness characteristics for a wide range of operation

Use of TCSR with Split Windings for Shortening the Spar Cycle Time in 500 kV Lines

Energy Technology Data Exchange (ETDEWEB)

Matinyan, A. M., E-mail: al-drm@mail.ru; Peshkov, M. V.; Karpov, V. N.; Alekseev, N. A. [JSC “R& D Center at Federal Grid Company of Unified Power System,” (Russian Federation)

2017-01-15

The arc-fault recharge phenomenon in single-phase automatic reclosure (SPAR) of a line is examined. Abrief description is given of the design of a 500 kV thyristor controlled shunt reactor (TCSR) with split valve-side windings. This type of TCSR is shown to effectively quench a single-phase arc fault in a power transmission line and shortens the SPAR cycle time.

Method and device for controlling nuclear reactor power

International Nuclear Information System (INIS)

Takigawa, Yukio; Ebata, Shigeo.

1988-01-01

Purpose: To detect and suppress the special power oscillations in the reactor core. Method: Four pairs of LPRM detectors, each pair comprising two detectors are disposed at an identical axial direction of the reactor core and situated at substantially insymmetrical positions at least in longitudinal, vertical and orthogonal directions with respect to the center of te reactor core and LPRM signals from them are inputted into a device for judging special power oscillations. In this case, a standardized mutual relation function is determined on every pair for the respective LPRM signals. Generation of special power oscillations in the reactor core is judged when it is detected that peaks appearing at least in one of the function forms for each pair are negative and have absolute values exceeding a predetermined value and that time of peak is within a predetermined time. The judged signal is inputted to a selected control rod insertion device. The selected control rod insertion device, upon preceiving the signal, inserts selected control rods into the reactor core to suppress the special power oscillations. Accordingly, it is possible to improve the fuel integrity. (Horiuchi, T.)

Fuzzy power control algorithm for a pressurized water reactor

International Nuclear Information System (INIS)

Hah, Y.J.; Lee, B.W.

1994-01-01

A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations

Computer simulation system of neural PID control on nuclear reactor

International Nuclear Information System (INIS)

Chen Yuzhong; Yang Kaijun; Shen Yongping

2001-01-01

Neural network proportional integral differential (PID) controller on nuclear reactor is designed, and the control process is simulated by computer. The simulation result show that neutral network PID controller can automatically adjust its parameter to ideal state, and good control result can be gotten in reactor control process

On-line computer control of a nuclear reactor using optimal control and state estimation methods

International Nuclear Information System (INIS)

Tye, C.

1980-01-01

This paper describes the experimental implementation of a nuclear reactor control system using combined optimal state feedback based on the Quadratic Regulator and state estimation using Kalman filtering techniques. The results obtained from the experiments indicate that a reactor control loop designed using this approach has improved stability margins, greater speed of response and noise filtering properties compared with a conventional reactor control loop. 11 refs

Problems of control of WWER-type pressurized water reactors (PWR's)

International Nuclear Information System (INIS)

Drab, F.; Grof, V.

1978-01-01

The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

Control rod studies in small and medium sized fast reactors

International Nuclear Information System (INIS)

John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

1988-01-01

Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

Overall aspects of control of ISIS-type nuclear reactor

International Nuclear Information System (INIS)

Amato, S.; Santinelli, A.

1996-01-01

The paper describes the main aspects related to the definition of main controls required to operate an ISIS-type nuclear power reactors. ISIS is a PWR-type intrinsically safe nuclear reactor designed by ANSALDO, based on density lock concept; it presents, between the other safety functions, self-depressurization and core cooling capability for unlimited time. Due to its specific characteristics, the ISIS reactor required to development of new control philosophy (if compared with actual nuclear power reactor) with the implementation of new control functions, for instance the density locks hot/cold interface locations control. This paper describes the main control functions implemented, their rationale, as well as the dynamic simulation performed to verify the adequacy of controls definitions. The dynamic simulations here described refers to a step-wise power ramp of 100-90-100 (% of nominal power) and to a power ramp of 100-50-100 with a slope of 5%/min; the results obtained have shown the ISIS capability to perform such operational transients, despite its innovative design was mainly focused on intrinsically safe behaviour. (author)

Method for controlling FBR type reactor

International Nuclear Information System (INIS)

Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

1991-01-01

The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

Method and apparatus for controlling the neutron flux in nuclear reactors

International Nuclear Information System (INIS)

Minnick, L.E.

1979-01-01

A control rod assembly in a nuclear reactor that automatically scrams the reactor when a loss of coolant flow occurs and that can also control the level of neutron flux in the reactor is described. The control rod assembly includes a separator plate having an orifice through which the reactor coolant flows and a sealing surface around the orifice. The control rod in the assembly has a complementary sealing surface. When the control rod and separator plate are brought into contact, the differential pressure across the separator plate caused by the flow of the primary coolant through the reactor core retains the two sealing surfaces together. If the flow of coolant stops or the differential pressure across the separator plate decreases for any reason, the control rod drops by gravity and the reactor is scrammed. The control rod is also automatically dropped as a result of the lateral vibration of an earthquake or by the downward motion of the rod drive shaft, either of which will open the sealing surfaces and reduce the sealing pressure

A digital controller for the Omega West Reactor

International Nuclear Information System (INIS)

Minor, M.M.; Kaufman, M.D.; Smith, T.W.

1992-05-01

A new nuclear reactor control system for the Omega West Reactor (OWR) has been designed to replace the aging and hard to maintain controller presently installed. The controller uses single board computers, digital and analog input and output modules, and stepping motor indexers installed on a standard bus (VME bus). The eight poison control rod drive motors are replaced with stepping motors. The control algorithm for the OWR was not changed in order to expedite approval for installation. This report presents the results of the development of the new control system. Included in the report are copies of some of the software that drives the new controller

Water quality control device and water quality control method for reactor primary coolant system

International Nuclear Information System (INIS)

Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

1995-01-01

The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

Characterisation of reactor control rod drives. Specification 1-6

International Nuclear Information System (INIS)

1975-03-01

The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

Analysis of man-machine interaction for control and display system in main control room of light water reactor

International Nuclear Information System (INIS)

Santosa, Kussigit; Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Darlis; Sudiono, Bambang

1998-01-01

One of potential hazard in Nuclear Power Plant is the failure of its operation. The accident or operation failure in the reactor must be concerned event its probability is low. The important thing should be concerned is 'Analysis of Man-Machine Interaction (MMI) for Control and Display System in Main Control Room (MCR) of Nuclear Power Reactor', especially LWR type. Control and Display System in MCR of Reactor is the main part of MMI link process in Reactor MCR work system. Signal from display system showed performance process in reactor, while this signal will be received by operator. This signal will be described through central nerve for making decision what kind must be done. Then the operator manage the next process of reactor operation through control system. So by knowing Analysis of Man-Machine Interaction for Control and Display System in Main Control Room of Power Reactor, we can understand human error probability of the operator in reactor operation

Research Reactor Power Control System Design by MATLAB/SIMULINK

International Nuclear Information System (INIS)

Baang, Dane; Suh, Yong Suk; Kim, Young Ki; Im, Ki Hong

2013-01-01

In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

Enabling autonomous control for space reactor power systems

International Nuclear Information System (INIS)

Wood, R. T.

2006-01-01

The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Reactor control device for controlling load of nuclear power plant

International Nuclear Information System (INIS)

Hirota, Tadakuni; Yokoyama, Terukuni; Masuda, Jiro.

1981-01-01

Purpose: To improve the load follow-up capacity of a nuclear reactor by automatically controlling the width of the not-sensing band of a control rod inserting and removing discriminator circuit. Constitution: When load control operations such as automatic load control, automatic frequency control, governor free operation and so forth are conducted, the width of a not sensing band of a control rod inserting and removing discriminator circuit is ao automatically controlled that the not sensing band width may return to ordinary value in a normal operation by avoiding the fast repetition of inserting and removing control rods by increasing the width of the insensing band if the period of a control deviation signal produced due to the variation in the load is quickly repeated and varied in correspondence to the control deviation signal. That is, a circuit for varying the insensing band of the control circuit for driving a control mechanism is provided to reduce the amount of driving the control rods in a load control operation and to reduce the strain of the power distribution of the nuclear reactor, thereby improving the load control capacity. (Yoshihara, H.)

Methods for reactor physics calculations for control rods in fast reactors

International Nuclear Information System (INIS)

Grimstone, M.J.; Rowlands, J.L.

1988-12-01

The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

Innovative Control concepts for German pressurized water reactors

International Nuclear Information System (INIS)

Brzozowski, Raphael; Kuhn, Andreas

2010-01-01

Controlling reactor power without any manual support is becoming more and more important. The READIG project (READIG = Reactor Instrumentation and Digital Control) power control system installed in unit 2 of the Philippsburg nuclear power station (KKP 2) requires no manual intervention except for specific strategy criteria settings. It was even possible to eliminate the power distribution set points. With minor adaptations, this concept can be applied in other PWR plants as well. KKP 2 is a PWR plant with particularly sophisticated core charges; as a consequence, the I and C systems were adapted accordingly. The increase in integral reactor power and the low-leakage core charges are the main reasons for lower limiting margins, especially in peak limiting. The standard control concept was supplemented in such a way that a more precise fine control concept for power distribution in the full-load regime is achieved. The READIG project fully utilizes the possibilities offered by digital TXS Technology, which is why use is also made of physical parameterization. The new power distribution control concept has these advantages: - Operation at small peak-/DNB-reactor output limitation margins. - Stable control without manual intervention also in load cycles and in the frequency control mode. - Simplified operation due to omission of the power distribution set point. - Reduction to zero of the frequency of L-bank steps at constant power with superimposed frequency control mode. - Reduction to zero of the frequency of D-bank steps at constant power with superimposed frequency control mode. - Lower quantities of demineralized water to be fed at constant power with superimposed frequency control mode (±1%). (orig.)

Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Borio Di Tigliole, A.; Magrotti, G. [Laboratorio Energia Nucleare Applicata (L.E.N.A.), University of Pavia, Via Aselli 41, 27100 (Italy); Cammi, A.; Memoli, V. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division (CeSNEF), Via Ponzio 34/3, 20133 Milano (Italy); Gadan, M. A. [Instrumentation and Control Department, National Atomic Energy Comission of Argentina, University of Pavia (Italy)

2009-07-01

The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

Reactor water chemistry control

International Nuclear Information System (INIS)

Kundu, A.K.

2010-01-01

Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2002-01-01

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

Modernization of turbine control system and reactor control system in Almaraz 1 and 2

International Nuclear Information System (INIS)

Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

2005-01-01

The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

Multivariable Feedback Control of Nuclear Reactors

Directory of Open Access Journals (Sweden)

Rune Moen

1982-07-01

Full Text Available Multivariable feedback control has been adapted for optimal control of the spatial power distribution in nuclear reactor cores. Two design techniques, based on the theory of automatic control, were developed: the State Variable Feedback (SVF is an application of the linear optimal control theory, and the Multivariable Frequency Response (MFR is based on a generalization of the traditional frequency response approach to control system design.

On some control problems of dynamic of reactor

Science.gov (United States)

Baskakov, A. V.; Volkov, N. P.

2017-12-01

The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

Control rod for nuclear reactor

International Nuclear Information System (INIS)

Tada, Kaoru; Kawano, Shohei

1998-01-01

A guide roller is prepared by forming an oxide membrane on the surface of a molded roller product comprising, as a material, a deposition-reinforced type nickel-based alloy reinforced by deposition of fine particles by applying a heat treatment to a nickel-based alloy. When the guide roller is used in reactor water, since the roller has an oxide membrane on the surface, leaching of nickel to reactor water is reduced, and radioactive corrosive products including cobalt 58 are reduced to decrease an operator's exposure dose upon periodical inspections of a plant. The oxide membrane is formed by applying heat treatment under an oxidative atmosphere. Then, the amount of abrasion of pins and rollers in association with start-up or shut down of a reactor and control of the power can be reduced thereby enabling to suppress increase of radiation dose due to cobalt 60 and cobalt 58. (N.H.)

Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

International Nuclear Information System (INIS)

Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

2013-01-01

RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

Fuzzy algorithm for an automatic reactor power control in a PWR

International Nuclear Information System (INIS)

Hah, Yung Joon; Song, In Ho; Yu, Sung Sik; Choi, Jung In; Lee, Byong Whi

1994-01-01

A fuzzy algorithm is presented for automatic reactor power control in a pressurized water reactor. Automatic power shape control is complicated by the use of control rods because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability for the load - follow operation including frequency control. In an attempt to achieve automatic power shape control without any design modification of the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multi - input multi - output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to the Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of the pressurized water reactor during the load - follow operation

Reactor power automatically controlling method and device for BWR type reactor

International Nuclear Information System (INIS)

Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

1997-01-01

For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

Use of reactivity constraints for the automatic control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1985-01-01

A theoretical framework for the automatic control of reactor power has been developed and experimentally evaluated on the 5 MWt Research Reactor that is operated by the Massachusetts Institute of Technology. The controller functions by restricting the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally designated as ''feasibility of control''. It has been shown experimentally that maintenance of feasibility of control is a sufficient condition for the automatic control of reactor power. This research should be of value in the design of closed-loop controllers, in the creation of reactivity displays, in the provision of guidance to operators regarding the timing of reactivity changes, and as an experimental envelope within which alternate control strategies can be evaluated

The matter of probability controlling melting of nuclear ship reactor

International Nuclear Information System (INIS)

Pihowicz, W.; Sobczyk, S.

2008-01-01

In the first part of this work beside description of split power, power of radioactivity disintegration and afterpower and its ability to extinguish, the genera condition of melting nuclear reactor core and its detailed versions were described. This paper also include the description of consequences melting nuclear reactor core both in case of stationary and mobile (ship) reactor and underline substantial differences. Next, fulfilled with succeed, control under melting of stationary nuclear reactor core was characterized.The middle part describe author's idea of controlling melting of nuclear ship reactor core. It is based on: - the suggestion of prevention pressure's untightness in safety tank of nuclear ship reactor by '' corium '' - and the suggestion of preventing walls of this tank from melting by '' corium ''. In the end the technological and construction barriers of the prevention from melting nuclear ship reactor and draw conclusions was presented. (author)

Integrated Management System, Configuration and Document Control for Research Reactors

International Nuclear Information System (INIS)

Steynberg, B.J.; Bruyn, J.F. du

2017-01-01

An integrated management system is a single management framework establishing all the processes necessary for the organisation to address all its goals and objectives. Very often only quality, environment and health & safety goals are included when referred to an integrated management system. However, within the research reactor environment such system should include goals pertinent to economic, environmental, health, operational, quality, safeguards, safety, security, and social considerations. One of the important objectives of an integrated management is to create the environment for a healthy safety culture. Configuration management is a disciplined process that involves both management and technical direction to establish and document the design requirements and the physical configuration of the research reactor and to ensure that they remain consistent with each other and the documentation. Configuration is the combination of the physical, functional, and operational characteristics of the structures, systems, and components (SSCs) or parts of the research reactor, operation, or activity. The basic objectives and general principles of configuration management are the same for all research reactors. The objectives of configuration management are to: a) Establish consistency among design requirements, physical configuration, and documentation (including analyses, drawings, and procedures) for the research reactor; b) Maintain this consistency throughout the life of the research reactor, particularly as changes are being made; and c) Retain confidence in the safety of the research reactor. The key elements needed to manage the configuration of research reactors are design requirements, work control, change control, document control, and configuration management assessments. The objective of document control is to ensure that only the most recently approved versions of documents are used in the process of operating, maintaining, and modifying the research reactor

Experimental development of power reactor intelligent control

International Nuclear Information System (INIS)

Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

1992-01-01

The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

Autonomous Control of Space Reactor Systems

International Nuclear Information System (INIS)

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

2007-01-01

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are available to perform intelligent control functions that are necessary for both normal and abnormal operational conditions

Autonomous Control of Space Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

2007-11-30

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

Rolls-Royce digital Rod Control System

International Nuclear Information System (INIS)

Pouillot, M.

2010-01-01

Full text of publication follows: Rolls-Royce has developed a new generation of Rod Control System, based on 40 years of experience. The fifth-generation Rod Control System (RCS) from Rolls-Royce offers a reliable, modular design with adaptability to your preferred platform, for modernization projects or new reactors. Flexible implementation provides the option for you to keep existing cabinets, which permits you to optimize installation approach. Main features for the power part: - Control Rod Drive Mechanism (CRDM) type: 3-coil. - Independent control of each sub-bank. - Each sub-bank is controlled by a cycler unit and 3 identical power racks, each including 4 identical power modules and a common power-supply module. - Coil-per-coil digital control: each power module embeds power-conversion, current-control, and current-monitoring functions for one coil. Control and monitoring are carried out by separate electronics in the module. Current is digitized and fully monitored by means of min-max templates. - A double-hold function is included: a power module assigned to a gripper will activate its coil if a fault risking to cause a reactor trip occurs. - Power modules are standardized, hot-pluggable and self-configured: a power module includes a set of parameters for each type of coil SG, MG, LC. The module recognizes the rack it is plugged in, and chooses automatically parameters to be used. Main benefits: - Reduced operational, maintenance, training, and inventory costs: standardization of power modules and integration of control and monitoring on the same PC-card lead to a drastic reduction of spare part types, and simplification of the system. - Easy maintenance: - Replacement of a power module solves nearly all failures due to current control or monitoring for a coil. It is done instantly thanks to hot-plug capability. - On the front plate of power-modules, LEDs provide useful information for diagnostic: current setpoint from cycler, output current bar

Constrained model predictive control for load-following operation of APR reactors

International Nuclear Information System (INIS)

Kim, Jae Hwan; Lee, Sim Won; Kim, Ju Hyun; Na, Man Gyun; Yu, Keuk Jong; Kim, Han Gon

2012-01-01

The load-following operation of APR+ reactor is needed to control the power effectively using the control rods and to restrain the reactivity control from using the boric acid for flexibility of plant operation. Usually, the reason why the disproportion of axial flux distribution occurs during load-following operation is xenon-induced oscillation. The xenon has a very high absorption cross-section and makes the impact on the reactor delayed by the iodine precursor. The power maneuvering using automatically load-following operation has advantage in terms of safety and economic operation of the reactor, so the controller has to be designed efficiently. Therefore, an advanced control method that meets the conditions such as automatic control, flexibility, safety, and convenience is necessary to load-following operation of APR+ reactor. In this paper, the constrained model predictive control (MPC) method is applied to design APR reactor's automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control. Some controllers use only the current tracking command, but MPC considers future commands in addition to the current tracking command. So, MPC can achieve better tracking performance than others. Furthermore, an MPC is to used in many industrial process control systems. The basic concept of the MPC is to solve an optimization problem for a finite future time interval at present time and to implement the first optimal control input as the current control input. The KISPAC-1D code, which models the APR+ nuclear power plants, is interfaced to the proposed controller to verify the tracking performance of the reactor power level and ASI. It is known that the proposed controller exhibits very fast tracking responses

Evaluation of 'period-generated' control laws for the time-optimal control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.

1988-01-01

Time-Optimal control of neutronic power has recently been achieved by developing control laws that determine the actuator mechanism velocity necessary to produce a specified reactor period. These laws are designated as the 'MIT-SNL Period-Generated Minimum Time Control Laws'. Relative to time-optimal response, they function by altering the rate of change of reactivity so that the instantaneous period is stepped from infinity to its minimum allowed value, held at that value until the desired power level is attained, and then stepped back to infinity. The results of a systematic evaluation of these laws are presented. The behavior of each term in the control laws is shown and the capability of these laws to control properly the reactor power is demonstrated. Factors affecting the implementation of these laws, such as the prompt neutron lifetime and the differential reactivity worth of the actuators, are discussed. Finally, the results of an experimental study in which these laws were used to adjust the power of the 5 MWt MIT Research Reactor are shown. The information presented should be of interest to those designing high performance control systems for test, spacecraft, or, in certain instances, commercial reactors

Reactivity control of nuclear power reactors: new options

International Nuclear Information System (INIS)

Alcala, F.

1984-01-01

Some actual aspects (referring to economy, non-proliferation and environmental impact) of nuclear power reactors has been analyzed from the point of view of the reactivity control physics. Specially studied have been the physical mechanisms related with the spectral shift control method and their general positive effects on those aspects. The analysis carried out suggested the application of the above method of control to reactors with non-hydrogenous fuel cells, which are mainly characterized by their high moderator/fuel ratio. Finally three different types of such fuel cells are presented and some results about one of them (belonging to a PHWR controlled by graphite rods) are given. (author)

Development of automated controller system for controlling reactivity by using FPGA in research reactor application

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib

2012-01-01

The scope for this research paper is to produce a detail design for Development of Automated Controller System for Controlling Reactivity by using FPGA in Research Reactor Application for high safety nuclear operation. The development of this project including design, purchasing, fabrication, installation, testing and validation and verification for one prototype automated controller system for controlling reactivity in industry local technology for human capacity and capability development towards the first Nuclear Power Programme (NPP) in Malaysia. The specific objectives of this research paper are to Development of Automated Controller System for Controlling Reactivity (ACSCR) in Research Reactor Application (PUSPATI TRIGA Reactor) by using simultaneous movement method; To design, fabricate and produce the accuracy of Control Rods Drive Mechanism to 0.1 mm resolution using a stepper motor as an actuator; To design, install and produce the system response to be more faster by using Field Programmable Gate Array (FPGA) and High Speed Computer; and to improve the Safety Level of the Research Reactor in high safety nuclear operation condition. (author)

molecular weight control of a batch suspension polymerization reactor

International Nuclear Information System (INIS)

Shahrokhi, M.; Fanaei, M. A.

2002-01-01

This paper concerns molecular weight control of a batch polymerization reactor where suspension polymerization of methyl methylacrylate (MMA) takes place. For this purpose, a cascade control structure with two control loops has been selected. The slave loop is used for temperature control using on-line temperature measurements, and the master loop controls the average molecular weights based on its estimated values. Two different control algorithms namely proportional-integral (PI) controller and globally linearizing controller (GLC) have been used for temperature control. An estimator, which has the structure of an extended Kalman filter(EKF), is used for estimating monomer conversion and average molecular weights of polymer using reactor temperature measurements. The performance of proposed control algorithm is evaluated through simulation and experimental studies. The results indicate that a constant average molecular weight cannot be achieved in case of strong gel effect. However, the polydispersity of product will be lower in comparison to isothermal operation. It is also shown that in case of mo dek mismatch, the performance of cascade control is superior compared to the case where only reactor temperature is controlled based on desired temperature trajectory obtained through cascade strategy

Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2003-01-01

The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

Control of ZrH reactor reactivity perturbations during orbital maneuvers

International Nuclear Information System (INIS)

Audette, R.F.

1970-01-01

Scheduled and inadvertent vehicle maneuvers in manned and unmanned space missions may result in reactivity perturbations to the ZrH reactor due to fuel and control drum motion from acceleration forces. Potential power and outlet coolant temperature excursions could result in interruptions of PCS power generation, or excessive coolant temperatures if uncontrolled. This analysis compares potential uncontrolled reactor transients with allowable transients for uninterrupted electrical power generation from a Brayton system, and presents a control scheme to limit transient reactor outlet temperatures to 1250 0 F for a system designed to operate at a nominal 1200 0 F reactor outlet. Potential uncontrolled transients could result in a reactor outlet temperature swing of +-77 0 F about a nominal 1200 0 F and a reactor power swing of +92 Kwt and -67 Kwt about a nominal 130 Kwt for the Brayton System. (U.S.)

Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

Energy Technology Data Exchange (ETDEWEB)

Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

2005-07-01

The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor; Control difuso adaptable de la potencia neutronica del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Rojas R, E.

2014-07-01

The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n{sub o} up to a predetermined final level n{sub f}. During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

A basic design of SR4 instrumentation and control system for research reactor

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

Application of 2DOF controller for reactor power control. Verification by numerical simulation

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Suzuki, Katsuo

1996-09-01

In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

International Nuclear Information System (INIS)

Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

Energy Technology Data Exchange (ETDEWEB)

Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Dept. of Nuclear Engineering)

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs.

Cyber security for remote monitoring and control of small reactors

Energy Technology Data Exchange (ETDEWEB)

Trask, D., E-mail: dave.trask@cnl.ca [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Jung, C. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); MacDonald, M., E-mail: marienna.macdonald@cnl.ca [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

There is growing international interest and activity in the development of small nuclear reactor technology with a number of vendors interested in building small reactors in Canada to serve remote locations. A common theme of small reactor designs proposed for remote Canadian locations is the concept of a centrally located main control centre operating several remotely located reactors via satellite communications. This theme was echoed at a recent IAEA conference where a recommendation was made to study I&C for remotely controlled small modular reactors, including satellite links and cyber security. This paper summarizes the results of an AECL-CNSC research project to analyze satellite communication technologies used for remote monitoring and control functions in order to provide cyber security regulatory considerations. The scope of this research included a basic survey of existing satellite communications technology and its use in industrial control applications, a brief history of satellite vulnerabilities and a broad review of over 50 standards, guidelines, and regulations from recognized institutions covering safety, cyber security, and industrial communication networks including wireless communications in general. This paper concludes that satellite communications should not be arbitrarily excluded by standards or regulation from use for the remote control and monitoring of small nuclear reactors. Instead, reliance should be placed on processes that are independent of any particular technology, such as reducing risks by applying control measures and demonstrating required reliability through good design practices and testing. Ultimately, it is compliance to well-developed standards that yields the evidence to conclude whether a particular application that uses satellite communications is safe and secure. (author)

Cyber security for remote monitoring and control of small reactors

International Nuclear Information System (INIS)

Trask, D.; Jung, C.; MacDonald, M.

2014-01-01

There is growing international interest and activity in the development of small nuclear reactor technology with a number of vendors interested in building small reactors in Canada to serve remote locations. A common theme of small reactor designs proposed for remote Canadian locations is the concept of a centrally located main control centre operating several remotely located reactors via satellite communications. This theme was echoed at a recent IAEA conference where a recommendation was made to study I&C for remotely controlled small modular reactors, including satellite links and cyber security. This paper summarizes the results of an AECL-CNSC research project to analyze satellite communication technologies used for remote monitoring and control functions in order to provide cyber security regulatory considerations. The scope of this research included a basic survey of existing satellite communications technology and its use in industrial control applications, a brief history of satellite vulnerabilities and a broad review of over 50 standards, guidelines, and regulations from recognized institutions covering safety, cyber security, and industrial communication networks including wireless communications in general. This paper concludes that satellite communications should not be arbitrarily excluded by standards or regulation from use for the remote control and monitoring of small nuclear reactors. Instead, reliance should be placed on processes that are independent of any particular technology, such as reducing risks by applying control measures and demonstrating required reliability through good design practices and testing. Ultimately, it is compliance to well-developed standards that yields the evidence to conclude whether a particular application that uses satellite communications is safe and secure. (author)

Dimensional control and check of field machining parts for reactor internals installation

International Nuclear Information System (INIS)

Zhang Caifang

2010-01-01

Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

Self operation type reactor control device

International Nuclear Information System (INIS)

Saito, Makoto; Gunji, Minoru.

1990-01-01

A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

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

International Nuclear Information System (INIS)

Liu Cheng; Peng Jinfeng; Zhao Fuyu; Li Chong

2009-01-01

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.

Distributed computer control system for reactor optimization

International Nuclear Information System (INIS)

Williams, A.H.

1983-01-01

At the Oldbury power station a prototype distributed computer control system has been installed. This system is designed to support research and development into improved reactor temperature control methods. This work will lead to the development and demonstration of new optimal control systems for improvement of plant efficiency and increase of generated output. The system can collect plant data from special test instrumentation connected to dedicated scanners and from the station's existing data processing system. The system can also, via distributed microprocessor-based interface units, make adjustments to the desired reactor channel gas exit temperatures. The existing control equipment will then adjust the height of control rods to maintain operation at these temperatures. The design of the distributed system is based on extensive experience with distributed systems for direct digital control, operator display and plant monitoring. The paper describes various aspects of this system, with particular emphasis on: (1) the hierarchal system structure; (2) the modular construction of the system to facilitate installation, commissioning and testing, and to reduce maintenance to module replacement; (3) the integration of the system into the station's existing data processing system; (4) distributed microprocessor-based interfaces to the reactor controls, with extensive security facilities implemented by hardware and software; (5) data transfer using point-to-point and bussed data links; (6) man-machine communication based on VDUs with computer input push-buttons and touch-sensitive screens; and (7) the use of a software system supporting a high-level engineer-orientated programming language, at all levels in the system, together with comprehensive data link management

Method of nuclear reactor control using a variable temperature load dependent set point

International Nuclear Information System (INIS)

Kelly, J.J.; Rambo, G.E.

1982-01-01

A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

Nuclear reactor control with fuzzy logic approaches - strengths, weakness, opportunities, and threats

International Nuclear Information System (INIS)

Ruan, Da

2004-01-01

As part of the special track on 'Lessons learned from computational intelligence in nuclear applications' at the forthcoming FLINS 2004 conference on Applied Computational Intelligence (Blankenberge, Belgium, September 1-3, 2004), research experiences on fuzzy logic techniques in applications of nuclear reactor control operation are critically reviewed in this presentation. Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined thought a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK-CEN) and the Mexican Nuclear Centre (ININ) on the fuzzy logic control for nuclear reactor control project under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (Author)

Simulation and control of the site-dependent neutron density in a nuclear reactor

International Nuclear Information System (INIS)

Stark, K.

1974-01-01

The present work deals with the simulation and control of a pressurized-water reactor such as is used in nuclear power plants today. In the first part of the work, the mathematical model equations of the reactor are set up. They take into consideration the local distribution of the various reactor parameters as far as seems necessary for further investigations. Taking the given approximations, the mathematical model is locally one-dimensional; it is valid for the period of time in which a power control of the reactor must work. The model equations set up are calculated on an analog/hybride computer according to the modal simulation method in true time. The method is distinguished in the present problem here through good convergence and enables the observation of the simulation results as a stationary picture on an oscillograph screen. For this reason, a simulation of this type seems particularly suitable for the training of operational personnel. The aim of the second part of the work is the development of a simple control concept which enables the control of the total power of the reactor as well as of the distribution of the power density in the reactor core. The fundamentals of the control design are the non-linear system equations of the nuclear reactor. The developed control is based on the controlling of eigenfunctions; it controls the total power of the reactor as well as the distribution of the power density in the reactor core where a uniform burn-up of the nuclear fuel is seen to. Part-absorbing control rods amongst others are used as actuators like they are already used in that type of reactors. (orig./LH) [de

Autonomous Control Capabilities for Space Reactor Power Systems

International Nuclear Information System (INIS)

Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

2004-01-01

The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission

Device for controlling water supply to nuclear reactor

International Nuclear Information System (INIS)

Iwasaki, Toshio.

1974-01-01

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

A nuclear reactor power controller using a receding horizon control method

International Nuclear Information System (INIS)

Na, Man Gyun; Sim, Young Rok

2001-01-01

A receding horizon control method is applied to design a fully automatic controller for thermal power in a reactor core. The basic concept of the receding horizon control is to solve an optimization problem for a finite future at current time and to implement as the current control input the first optimal control input among the solutions of the finite time steps. The procedure is then repeated at each subsequent instant. The receding horizon controller is designed so that the difference between the output and the desired output is minimized and the variation of the control rod position is small. The nonlinear PWR plant model (nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) was used to verify the proposed controller of reactor power. And a controller design model used for designing the receding horizon controller was obtained by applying a parameter estimation algorithm. From numerical simulation results, the performances of this controller for the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are design requirements are proved to be excellent

Design for the human-machine interface of a digitalized reactor control-room

International Nuclear Information System (INIS)

Qu Ronghong; Zhang Liangju; Li Duo; Yu Hui

2005-01-01

Digitalized technology is implemented in the instrumentation and control system of an in-construction research reactor, which advances information display in both contents and styles in a nuclear reactor control-room, and greatly improves human-machine interface. In the design for a digitalized nuclear reactor control-room there are a series of new problems and technologies should be considered seriously. This paper mainly introduces the design for the digitalized control-room of the research nuclear reactor and covered topics include design principle of human-machine interface, organization and classification of interface graphics, technologies and principles based on human factors engineering and implemented in the graphics design. (authors)

Power control device for heavy water moderated reactor

International Nuclear Information System (INIS)

Matsushima, Hidesuke; Masuda, Hiroyuki.

1978-01-01

Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Reactor trip on turbine trip inhibit control system for nuclear power generating system

International Nuclear Information System (INIS)

Torres, J.M.; Musick, C.R.

1976-01-01

A reactor trip on turbine trip inhibit control system for a nuclear power generating system which utilizes steam bypass valves is described. The control system inhibits a normally automatic reactor trip on turbine trip when the bypass valves have the capability of bypassing enough steam to prevent reactor trip limits from being reached and/or to prevent opening of the secondary safety pressure valves. The control system generates a bypass valve capability signal which is continuously compared with the reactor power. If the capability is greater than the reactor power, then an inhibit signal is generated which prevents a turbine trip signal from tripping the nuclear reactor. 10 claims, 4 figures

Design analysis and microprocessor based control of a nuclear reactor

International Nuclear Information System (INIS)

Sabbakh, N.J.

1988-01-01

The object of this thesis is to design and test a microprocessor based controller, to a simulated nuclear reactor system. The mathematical model that describes the dynamics of a typical nuclear reactor of one group of delayed neutrons approximations with temperature feedback was chosen. A digital computer program has been developed for the design and analysis of a simulated model based on the concept of state-variable feedback in order to meet a desired system response with maximum overshoot of 3.4% and setting time of 4 sec. The state variable feedback coefficients are designed for the continuous system, then an approximation is used to obtain in the state variable feedback vector for the discrete system. System control was implemented utilizing Direct Digital Control (DDC) of a nuclear reactor simulated model through a control algorithm that was performed by means of a microprocessor based system. The controller performance was satisfactorily tested by exciting the reactor system with a transient reactivity disturbance and by a step change in power demand. Direct digital control, when implemented on a microprocessor adds versatility, flexibility in system design with the added advantage of possible use of optimal control algorithms. 6 tabs.; 30 figs.; 46 refs.; 6 apps

Control unit of a nuclear reactor

International Nuclear Information System (INIS)

Desfontaines, Guy; Le Helloco, Michel.

1981-01-01

Control unit comprising multiple leak-tight vessels, in communication with the inside of the reactor vessel, extending this vessel above its cover, in the vertical direction and each one enclosing a mechanism for moving a cluster of material absorbing the neutrons in the reactor core, actuated by a motor. This control unit is of reduced height above the vessel cover and provides efficient protection of the leak tight containments and the mechanisms in the event of earthquakes, easy removal and refitting of the vessel cover, good ventilation of the power devices of the mechanisms without the use of a complex ventilation system, efficient thermal insulation of the leak-tight containments assembly, as well as easy access to the motors and mechanism located in the leak-tight containment for carrying out any maintenance and repairs that might be required [fr

Control of operational transients in power reactors - Methodology

International Nuclear Information System (INIS)

Vukovic, D.

1983-01-01

By introducing the nuclear power stations in the electric power system, questions of their possibilities to satisfy system's demand arise. Control of operational transients (temperature and Xe 135 ) in power reactors by determining the optimal control rod strategy is given. Ti optimize the Xe 135 transients, the Pantryagin theorem of optimal processes is applied. For solving three dimensional, two-group diffusion equations the heterogeneous Feinberg-Galanin method with axial flux harmonics is adopted. An application of this formalism to three-dimensional, finite cylindrical pressurised water reactor radially reflected is presented. (author)

Modelling of Control Bars in Calculations of Boiling Water Reactors

International Nuclear Information System (INIS)

Khlaifi, A.; Buiron, L.

2004-01-01

The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

Cobalt-60 control in Ontario Hydro reactors

International Nuclear Information System (INIS)

Lacy, C.S.

1988-01-01

This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

International Nuclear Information System (INIS)

Ozaki, Yoshihiko; Sunagawa, Takeyoshi

2014-01-01

In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

Neutron field control cybernetics model of RBMK reactor operator

International Nuclear Information System (INIS)

Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

1992-01-01

Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

Simulation of the TREAT-Upgrade Automatic Reactor Control System

International Nuclear Information System (INIS)

Lipinski, W.C.; Kirsch, L.W.; Valente, A.D.

1984-01-01

This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility

Research reactor instrumentation and control technology. Report of a technical committee meeting

International Nuclear Information System (INIS)

1997-10-01

The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

Operation control equipment for BWR type reactor

International Nuclear Information System (INIS)

Izumi, Masayuki; Takeda, Renzo.

1981-01-01

Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

Use of university research reactors to teach control engineering

International Nuclear Information System (INIS)

Bernard, J.A.

1991-01-01

University research reactors (URRs) have provided generations of students with the opportunity to receive instruction and do hands-on work in reactor dynamics, neutron scattering, health physics, and neutron activation analysis. Given that many URRs are currently converting to programmable control systems, the opportunity now exists to provide a similar learning experience to those studying systems control engineering. That possibility is examined here with emphasis on the need for the inclusion of experiment in control engineering curricula, the type of activities that could be performed, and safety considerations

Globally linearized control on diabatic continuous stirred tank reactor: a case study.

Science.gov (United States)

Jana, Amiya Kumar; Samanta, Amar Nath; Ganguly, Saibal

2005-07-01

This paper focuses on the promise of globally linearized control (GLC) structure in the realm of strongly nonlinear reactor system control. The proposed nonlinear control strategy is comprised of: (i) an input-output linearizing state feedback law (transformer), (ii) a state observer, and (iii) an external linear controller. The synthesis of discrete-time GLC controller for single-input single-output diabatic continuous stirred tank reactor (DCSTR) has been studied first, followed by the synthesis of feedforward/feedback controller for the same reactor having dead time in process as well as in disturbance. Subsequently, the multivariable GLC structure has been designed and then applied on multi-input multi-output DCSTR system. The simulation study shows high quality performance of the derived nonlinear controllers. The better-performed GLC in conjunction with reduced-order observer has been compared with the conventional proportional integral controller on the example reactor and superior performance has been achieved by the proposed GLC control scheme.

Inherent reactor power controller for a metal-fueled ALMR

International Nuclear Information System (INIS)

Wood, R.T.; Wilson, T.L. Jr.

1990-01-01

Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

Cyclic movement pin mechanism for controlling a nuclear reactor

International Nuclear Information System (INIS)

Joly, J.G.; Martin, Jean.

1981-01-01

This invention concerns a recurring movement pin mechanism for controlling a nuclear reactor by shifting a neutron absorbing assembly, vertically mobile in the nuclear reactor, to adjust the power and for emergency shut-down. This mechanism ensures a continuous movement and accurate shut-down at any level of the travel height of the absorbing assembly in the core. It also prevents the impacts of the pivoting pins in the control rod slots [fr

Direct digital temperature control of the A-1 nuclear reactor

International Nuclear Information System (INIS)

Karpeta, C.

1975-01-01

The application is described of one of the modern control methods for designing an experimental digital temperature control system for heavy water moderated gas cooled reactors. The synthesis of the optimal stochastic regulator for reactor control in the area of the rated steady state was carried out using the method of dynamic programming and the Kalman filter technique. The analysis of the feedback circuit was conducted using control simulation on a universal digital computer. Results and experience are summed up. (author)

Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

International Nuclear Information System (INIS)

Goswami, S.

1983-01-01

The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

2003-07-01

Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

Control rod drives for HTGR type reactor

International Nuclear Information System (INIS)

Nishiguchi, Isoharu; Katagiri, Shigeo.

1991-01-01

The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

Control rod homogenization in heterogeneous sodium-cooled fast reactors

International Nuclear Information System (INIS)

Andersson, Mikael

2016-01-01

The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

Technique of nuclear reactors controls; Technique des controles des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Weill, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1953-12-15

This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [French] Nous avons aborde le probleme de la ''Technique du Controle des reacteurs nucleaires'' dans le but de realiser le controle du reacteur de Saclay. C'est ainsi que nous avons ete amene a etudier le probleme dans son ensemble, tel qu'il se pose pour tout reacteur a uranium naturel. Ce travail traite principalement du domaine des mesures a caractere nucleaire et s'etend dans le domaine des mesures thermodynamque de niveaux, etc... mettant en relief les caracteristiques nouvelles exigees de la part des detecteurs du fait de leur utilisation dans le flux de neutrons thermiques. Dans le domaine de mesures nucleaires, nous indiquons principalement les realisations et les resultats obtenus pour les detecteurs de neutrons thermiques et pour la mesure de courants d'ionisations. Nous traitons egalement du probleme technique du demarrage d'un reacteur et du probleme de la mesure de la reactivite. Nous donnons les details necessaires a la comrehension de tous les schemas et plans de cablages essentiels mis au point, en particulier, pour le reacteur de Saclay. (auteur)

Control of PWR reactor energy supplied to a stream turbine

International Nuclear Information System (INIS)

Petetrot, J.F.; Parent, Pierre.

1981-01-01

This patent presents a process for regulating the power provided by a pressurized water nuclear reactor to a steam turbine, by moving the control rods absorbing the neutrons in the reactor core and by diverting a fraction of the steam produced by the reactor, outside the turbine circuit, by opening by-pass valves [fr

Power distribution monitoring and control in the RBMK type reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Postnikov, V.V.; Volod'ko, Yu.I.

1980-01-01

Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

Automatic Control of Reactor Temperature and Power Distribution for a Daily Load following Operation

Energy Technology Data Exchange (ETDEWEB)

Yu, Keuk Jong; Kim, Han Gon [Korea Hydro and Nuclear Power Institute, Daejeon (Korea, Republic of)

2010-10-15

An automatic control method of reactor power and power distribution was developed for a daily load following operation of APR1400. This method used a model predictive control (MPC) methodology having second-order plant data. And it utilized a reactor power ratio and axial shape index as control variables. However, the reactor regulating system of APR1400 is operated by the difference between the average temperature of the reactor core and the reference temperature, which is proportional to the turbine load. Thus, this paper reports on the model predictive control methodology using fourth-order plant data and a reactor temperature instead of the reactor power shape. The purpose of this study is to develop a revised automatic controller and analyze the behavior of the nuclear reactor temperature (Tavg) and the axial shape index (ASI) using the MPC method during a daily load following operation

The slightly-enriched spectral shift control reactor

International Nuclear Information System (INIS)

Martin, W.R.; Lee, J.C.; Edlund, M.C.

1990-06-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in larger neutron captures in fertile 238 U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 show that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies have been carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, effort will focus on performing the final design calculations and continuing the research to develop improved methods for tight lattice analysis

State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

Energy Technology Data Exchange (ETDEWEB)

Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

2017-02-15

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Control device for start-up of reactor depressurization system

International Nuclear Information System (INIS)

Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

1998-01-01

The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

Comparative analysis of nuclear reactor control system designs

International Nuclear Information System (INIS)

Russcher, G.E.

1975-01-01

Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

Humidity control device in a reactor container

International Nuclear Information System (INIS)

Aizawa, Motohiro; Igarashi, Hiroo; Osumi, Katsumi; Kimura, Takashi.

1986-01-01

Purpose: To provide a device capable of maintaining the inside of a container under high humidity circumstantial conditions causing less atmospheric corrosions, in order to prevent injuries due to atmospheric corrosions to smaller diameter stainless steel pipeways in the reactor container. Constitution: Stress corrosion cracks (SCC) to the smaller diameter stainless steel pipeways are caused dependent on the relative humidity and it is effective as the countermeasure against SCC to maintain the relative humidity at a low level less than 30 % or high level greater than 60 %. Based on the above findings, a humidity control device is disposed so as to maintain the relative humidity for the atmosphere within a reactor core on a higher humidity region. The device is adapted such that recycling gas in the dry-well coolant circuit is passed through an orifice to atomize the water introduced from feedwater pipe and introduce into a reactor core or such that the recycling gases in the dry-well cooling circuit are bubbled into water to remove chlorine gas in the reactor container gas thereby increasing the humidity in the reactor container. (Kamimura, M.)

Reactor power control systems in nuclear power plants

International Nuclear Information System (INIS)

Nakajima, Kazuo.

1980-01-01

Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

Controllability studies for an advanced CANDU boiling light water reactor

International Nuclear Information System (INIS)

Lepp, R.M.; Hinds, H.W.

1976-12-01

Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

Development and testing of control rod drives for ship reactors

International Nuclear Information System (INIS)

Bruelheide, K.; Mundt, D.; Peters, C.-H.; Manthey, H.-J.

1978-01-01

The following paper deals with the development and testings of a new control rod drive design for marine reactors. Starting from the good operating experience with the advanced pressurized water reactor (FDR) of the NS OTTO HAHN a control rod drive system with an hermetically sealed drive principle was developed. A prototype control rod drive system was put through extensive tests and developed ready for standard production at the 'Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt'

Optimal power and distribution control for weakly-coupled-core reactor

International Nuclear Information System (INIS)

Oohori, Takahumi; Kaji, Ikuo

1977-01-01

A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

Multi-component controllers in reactor physics optimality analysis

International Nuclear Information System (INIS)

Aldemir, T.

1978-01-01

An algorithm is developed for the optimality analysis of thermal reactor assemblies with multi-component control vectors. The neutronics of the system under consideration is assumed to be described by the two-group diffusion equations and constraints are imposed upon the state and control variables. It is shown that if the problem is such that the differential and algebraic equations describing the system can be cast into a linear form via a change of variables, the optimal control components are piecewise constant functions and the global optimal controller can be determined by investigating the properties of the influence functions. Two specific problems are solved utilizing this approach. A thermal reactor consisting of fuel, burnable poison and moderator is found to yield maximal power when the assembly consists of two poison zones and the power density is constant throughout the assembly. It is shown that certain variational relations have to be considered to maintain the activeness of the system equations as differential constraints. The problem of determining the maximum initial breeding ratio for a thermal reactor is solved by treating the fertile and fissile material absorption densities as controllers. The optimal core configurations are found to consist of three fuel zones for a bare assembly and two fuel zones for a reflected assembly. The optimum fissile material density is determined to be inversely proportional to the thermal flux

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-Kyu, E-mail: power@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Kim, Kwangmin; Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of)

2015-11-15

Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

International Nuclear Information System (INIS)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-01-01

Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

Application of robust fuzzy control in power control of nuclear reactor

International Nuclear Information System (INIS)

Liu Lei; Luan Xiuchun; Jin Guangyuan; Yu Tao; Rao Su

2013-01-01

Robust-fuzzy controller based on T-S fuzzy model was designed for real-time controlling of nuclear reactor power and adapting to the load changing of power grid. Local controller was designed by means of state feedback technique, and the global controller was designed by parallel distributed compensation (PDC) method. The result of solving linear matrix inequalities (LMI) proves that this controller is stable. The simulation shows that the nuclear power can be well controlled in three typical conditions by this controller. (authors)

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Control of Canadian once-through direct cycle supercritical water-cooled reactors

International Nuclear Information System (INIS)

Sun, Peiwei; Wang, Baosheng; Zhang, Jianmin; Su, Guanghui

2015-01-01

Highlights: • Dynamic characteristics of Canadian SCWR are analyzed. • Hybrid feedforward and feedback control is adopted to deal with cross-coupling. • Gain scheduling control with smooth weight is applied to deal with nonlinearity. • It demonstrates through simulation that the control requirements are satisfied. - Abstract: Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system

Study of high-pressure cryogenic pumps with different methods of delivery control

International Nuclear Information System (INIS)

Braun, V.M.; Brailovskii, Y.L.; Pavlenko, Y.A.; Tsokalo, I.V.

1986-01-01

This paper describes new reciprocating pumps with smooth control of delivery in a running pump. Control is effected either by changing the length of the piston stroke or by changing the speed of the driving motor. The individual features of the two methods are described. In the first method (mechanical), delivery is controlled in the range 50 to 100%. A separate actuating mechanism is needed to connect the pump to an automatic control system. This method complicates the driving mechanism and increases the bulk and cost of production. In the second method of controlling the speed of the electric motor, an electric drive fitted with a frequency thyristor is used. AC induction motors series 4A working at current frequencies of 60 HZ are used. By this method, delivery control could be enhanced by 1.3 times. Comparative tests were made on pumps using the above methods of control. The tests demonstrated the possibilities of using the frequency thyristor converters. The complexity and high cost of EKT type drives is largely compensated by the convenience and simplicity of control in a wide range. The mechanical control is advantageous only in low-output units

Control of tritium permeation through fusion reactor strucural materials

International Nuclear Information System (INIS)

Maroni, V.A.

1978-01-01

The intention of this paper is to provide a brief synopsis of the status of understanding and technology pertaining to the dissolution and permeation of tritium in fusion reactor materials. The following sections of this paper attempt to develop a simple perspective for understanding the consequences of these phenomena and the nature of the technical methodology being contemplated to control their impact on fusion reactor operation. Considered in order are: (1) the occurrence of tritium in the fusion fuel cycle, (2) a set of tentative criteria to guide the analysis of tritium containment and control strategies, (3) the basic mechanisms by which tritium may be released from a fusion plant, and (4) the methods currently under development to control the permeation-related release mechanisms. To provide background and support for these considerations, existing solubility and permeation data for the hydrogen isotopes are compared and correlated under conditions to be expected in fusion reactor systems

Multilayer robust control for safety enhancement of reactor operations

International Nuclear Information System (INIS)

Edwards, R.M.; Lee, K.Y.; Ray, A.

1991-01-01

A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems

Graphics and control for in-reactor operations

International Nuclear Information System (INIS)

Smith, A.L.

1996-01-01

A wide range of manipulator systems has been developed to carry out remotely operated inspection, repair and maintenance tasks at the Magnox reactors in the United Kingdom. A key factor in the improvement of these systems in recent years has been the extensive use of computer graphics as a real-time aid to the manipulator operator. This is exemplified by the reactor pressure vessel inspection work at the Bradwell reactor which is described in detail. The graphics sub-system of the control system for the manipulator plays a unique and wide-ranging role. The 3D modelling and simulation capability of the IGRIP software has contributed to the conceptual design, detailed path planning, rehearsal support, public relations, real-time manipulator display, post inspection documentation and quality assurance. (UK)

Research on reactor power controller based on artificial immune P and PID cascade control technology

International Nuclear Information System (INIS)

Cheng Shouyu; Peng Minjun; Liu Xinkai

2014-01-01

The Reactor Power control system usually adopts the traditional PID controller, the traditional PID controller can meet the operating requirements, but the control effect is not very good. In order to improve this condition, the paper proposes an immune P and PID cascade controller which based the immune mechanism of B-cell co-operating with T-cell, the nuclear power controller based on artificial immune is less reported. In order to verify and validate the control strategy, the designed controller debugs with the full-scope real-time simulation system of nuclear power plants. The simulation results shows that the immune controller can effectively improve the dynamic operating characteristics of the reactor system, and the immune controller is superior to the traditional PID controller in control performance. (authors)

Advanced liquid metal reactor plant control system

International Nuclear Information System (INIS)

Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

1993-01-01

The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

Vent control device for nuclear reactor container

International Nuclear Information System (INIS)

Kubota, Ryuji.

1989-01-01

The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

Energy Technology Data Exchange (ETDEWEB)

Zaidabadinejad, Majid; Ansarifar, Gholam Reza [Isfahan Univ. (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-11-15

In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

International Nuclear Information System (INIS)

Zaidabadinejad, Majid; Ansarifar, Gholam Reza

2017-01-01

In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

Synthesis of relay control systems for nuclear reactors

International Nuclear Information System (INIS)

Postnikov, N.S.

1996-01-01

The problem on stabilizing an oscillatory-unstable reactor by a single-link relay system, the characteristics whereof have a dead zone and hysteresis loop, is considered. The methodology of synthesis of feedback law, providing for stochastic steady-state mode of reactor operation with the minimum frequency of control impact introduction is proposed. This methodology is applicable to general-type relay systems with arbitrary oscillatory-unstable objects. 6 refs., 5 figs

New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

Energy Technology Data Exchange (ETDEWEB)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

International Nuclear Information System (INIS)

Severin, V.P.

2007-01-01

The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

Study, design and evaluation of nuclear reactor computer control system

International Nuclear Information System (INIS)

Menacer, S.

1988-01-01

Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

Some safety considerations in laser-controlled thermonuclear reactors. Final report

International Nuclear Information System (INIS)

Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

1978-07-01

A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E., E-mail: thiago.brito86@yahoo.com.br, E-mail: cabol@ufpe.br, E-mail: wagner@unicap.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Universidade Catolica de Pernambuco (UNICAP), Recife, PE (Brazil). Centro de Ciencias e Tecnologia

2017-11-01

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

International Nuclear Information System (INIS)

Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E.; Universidade Catolica de Pernambuco

2017-01-01

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J.

1982-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

Development and design of control rod drive mechanisms for pressurized water reactors

International Nuclear Information System (INIS)

Leme, Francisco Louzano

2003-01-01

The Control Rod Drive Mechanisms (CRDM) for a Pressurized Water Reactor (PWR) are equipment, integrated to the reactor pressure vessel, incorporating mechanical and electrical components designed to move and position the control rods to guarantee the control of power and shutdown of the nuclear reactor, during normal operation, either in emergency or accidental situations. The type of CRDM used in PWR reactors, whose detailed individual description will be presented in this monograph are the Roller-Nut and Magnetic-Jack. The environment, where the CRDM performs its above presented operational functions, includes direct contact with the fluid used as coolant peculiar to the interior of the reactor, and its associated chemical characteristics, the radiation field next to the reactor core, and also the temperature and pressure in the reactor pressure vessel. So the importance of the CRDM design requirements related to its safety functions are emphasized. Finally, some aspects related to the mechanical and structural design of CRDM of a case study, considering the CRDM for a PWR from the experimental nuclear plant to be applied by CTMSP (Centro Tecnologico da Marinha em Sao Paulo), are pointed out. The design and development of these equipment (author)

The slightly-enriched spectral shift control reactor

Energy Technology Data Exchange (ETDEWEB)

Martin, W.R.; Lee, J.C.; Larsen, E.W. (Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering); Edlund, M.C. (Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (United States). Dept. of Mechanical and Nuclear Engineering)

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.

The slightly-enriched spectral shift control reactor

International Nuclear Information System (INIS)

Martin, W.R.; Lee, J.C.; Larsen, E.W.; Edlund, M.C.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile 238 U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC

A 34 MW, 120 MJ pulsed dc supply for the UK tokamak reactor ''DITE''

International Nuclear Information System (INIS)

Fry, M.G.J.

1978-01-01

A static rectifier set supplying as much as 120 MJ of energy at a peak power of 34 MW to the toroidal magnet coils in the DITE experiment is described in detail. The power supply is designed to meet the stringent requirements concerning the maximum admissible peak reactive power. The rectifier is divided into two series-connected sections, one with diode bridges providing a fixed voltage, the other with thyristors which may work in either the rectifier or line-commutated inverter mode. The rectifier transformers are preceded by mechanical line-voltage regulators of the patented ''Interstep'' types, which comprise dual-voltage 11/33 kV tapped autotransformers and tap selection switches. Four parallel-connected thyristor bridges are used, perfect current sharing being assured by independent control of firing circuits. The firing circuits consist of non-conventional pulse generators and pulse amplifiers. Trains of precisely timed firing pulses are produced by using the phase-lock loop technique and the TTL logic. An extremely high noise immunity is achieved. (J.U.)

The control of emissions from nuclear power reactors in Canada

International Nuclear Information System (INIS)

Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

1988-01-01

Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

Expert system driven fuzzy control application to power reactors

International Nuclear Information System (INIS)

Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

1990-01-01

For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a ''supervisory'' routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment

Instrumentation and control improvements at Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Christensen, L.J.; Planchon, H.P.

1993-01-01

The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

Reactor power control method upon accidents of electrical power system

International Nuclear Information System (INIS)

Hirose, Masao.

1983-01-01

Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

Backfitting in Rossendorf research reactor control and instrumentation system

International Nuclear Information System (INIS)

Klebau, J.; Seidler, S.

1985-01-01

The paper generally describes a decentralized Hierarchical Information System (HIS) which has been developed for backfitting in Rossendorf Research Reactor (RFR) control and instrumentation system. The RFR was put into operation in 1957 and reconstructed from 2 MW up to a thermal power of 10 MW at the end of the sixties. Backfitting is planned by use of an advanced computerized control system for the next years. Main tasks of HIS are: Processmonitoring, online-disturbance analysis, technical diagnosis, direct digital control and use of a special industrial robot for discharging of irradiated materials out of the reactor. Experiences obtained by HIS during a testperiod will be presented. (author)

Monte Carlo verification of control-rod worth for the Savannah River K reactor

International Nuclear Information System (INIS)

Mosteller, R.D.

1992-01-01

The Savannah River K Reactor is a heavy-water reactor that relies on control-rod movement to control its reactivity and power distribution during normal operations. It is necessary, therefore, to have an accurate estimate of the reactivity worth of its control rods in order to predict the behavior of the reactor. Westinghouse Savannah River Company (WSRC) uses the GLASS lattice-physics code to calculate few-group cross sections for fuel and control-rod assemblies in the K reactor. This paper compares the control-rod worth calculated by GLASS to that calculated by the MCNP Monte Carlo program. The GLASS calculations utilize its standard 37-group cross-section library, while the MCNP calculations employ continuous-energy isotopic cross-section libraries derived from ENDF/B-V. The MCNP calculations therefore combine the most rigorous analytical model and the most accurate cross sections currently available for thermal-reactor analysis. Consequently, the MCNP results comprise a computational benchmark against which the accuracy of the GLASS code can be evaluated

Computerized supervision and control system for movement at the RP-10 reactor control rods bank

International Nuclear Information System (INIS)

Padilla M, C.E.

1998-01-01

The project involves the use of a compatible microcomputer, Labwindows/CVI software, as well as National Instruments data acquisition cards AT-MIO16-E10 and PC-DIO96 to modify the sequence of movement of the reactor's rods and control them from a graphic interface in a computer's monitor. This graphic presentation is set as console of virtual instruments from where rod movement can be conducted. Normal rod movement, bank rod movement, and rod calibration have been considered. These experiences involve different logic of rod movements, which will determine movement sequence. Control of the automatic range of a current amplifier module was also considered. This module is know as 'automatic pilot amplifier' and given the strategic location of its detector (compensated ionizing camera) at the reactor's core, it delivers neutron flux current considered as reference to superficial neutron flux distribution at the reactor's core. Lecture and monitoring of this signal allows taking the reactor to a certain power, current of this signal is proportional to the power we want the reactor to reach. Advantages obtained with this system include the update of the control console, more uniform distribution of neutron flux, with lower and uniform burnup of nuclear fuel. (author)

Enhanced situation awareness and decision making for an intelligent reconfigurable reactor power controller

International Nuclear Information System (INIS)

Kenney, S.J.; Edwards, R.M.

1996-01-01

A Learning Automata based intelligent reconfigurable controller has been adapted for use as a reactor power controller to achieve improved reactor temperature performance. The intelligent reconfigurable controller is capable of enforcing either a classical or an optimal reactor power controller based on control performance feedback. Four control performance evaluation measures: dynamically estimated average quadratic temperature error, power, rod reactivity and rod reactivity rate were developed to provide feedback to the control decision component of the intelligent reconfigurable controller. Fuzzy Logic and Neural Network controllers have been studied for inclusion in the bank of controllers that form the intermediate level of an enhanced intelligent reconfigurable reactor power controller (IRRPC). The increased number of alternatives available to the supervisory level of the IRRPC requires enhanced situation awareness. Additional performance measures have been designed and a method for synthesizing them into a single indication of the overall performance of the currently enforced reactor power controller has been conceptualized. Modification of the reward/penalty scheme implemented in the existing IRRPC to increase the quality of the supervisory level decision process has been studied. The logogen model of human memory (Morton, 1969) and individual controller design information could be used to allocate reward to the most appropriate controller. Methods for allocating supervisory level attention were also studied with the goal of maximizing learning rate

Control console conceptual design for sheet type fuels of Triga Mark-II reactor

International Nuclear Information System (INIS)

Eko Priyono; Kurnia Wibowo; Anang Susanto

2016-01-01

The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

Method of controlling the water quality in nuclear reactors

International Nuclear Information System (INIS)

Ibe, Hidefumi.

1985-01-01

Purpose: To obtain a simple and reliable water quality calculation system and water quality control method based thereon for the entire primary coolant circuits in BWR type reactors. Method: In a method of controlling the water quality of the reactor water by injecting hydrogen into the primary coolant circuits of a nuclear reactor, by utilizing a first linear relationship established between the concentration of oxygen and hydrogen in the main steam system and the concentration of radiolysis products in the reactor core and separators and mixing plenum portions, each of the above-mentioned concentrations is calculated from the concentrations for hydrogen or oxygen. Further, by utilizing the first linear relationship established between the concentrations for the oxygen and hydrogen in the recycling system and the concentration of the radiolysis products in the system from the downcomer to the lower plenum portion, the above-mentioned concentration is calculated from the concentration for oxygen and hydrogen. Then, the hydrogen injection rate into the primary coolant system is determined such that the calculated value takes an aimed value. (Ikeda, J.)

Development of a nuclear reactor control system simulator using virtual instruments

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

2011-01-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

Development of a nuclear reactor control system simulator using virtual instruments

Energy Technology Data Exchange (ETDEWEB)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

Control of radioactive material transport in sodium-cooled reactors

International Nuclear Information System (INIS)

Brehm, W.F.

1980-03-01

The Radioactivity Control Technology (RCT) program was established by the Department of Energy to develop and demonstrate methods to control radionuclide transport to ex-core regions of sodium-cooled reactors. This radioactive material is contained within the reactor heat transport system with any release to the environment well below limits established by regulations. However, maintenance, repair, decontamination, and disposal operations potentially expose plant workers to radiation fields arising from radionuclides transported to primary system components. This paper deals with radioactive material generated and transported during steady-state operation, which remains after 24 Na decay. Potential release of radioactivity during postulated accident conditions is not discussed. The control methods for radionuclide transport, with emphasis on new information obtained since the last Environmental Control Symposium, are described. Development of control methods is an achievable goal

Nuclear reactor control assembly

International Nuclear Information System (INIS)

Negron, S.B.

1991-01-01

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

1976-05-01

The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

Control system design for a 100 MW(th) research reactor

International Nuclear Information System (INIS)

Seshadri, S.N.; Ranganath, M.V.; Singh, Manjit.

1983-01-01

This paper presents the computer simulation carried out to evolve a suitable analog controller for a 100 MW(th) heavy water moderated research reactor under construction at Trombay. The control action is based on the average neutron flux in the reactor core and the reactivity is controlled by adjusting the moderator level in the calandria. A dual control scheme controlling the inflow as well as the outflow was adopted in order to fully exploit the capabilities of control elements. For reasons of reliability, the system consists of three identical channels enabling safe operation even under one channel failure. Based on the simulation studies a suitable compensation network was incorporated to achieve satisfactory system response. (author)

Conceptional design of the vertical field control system in JIPP T-II

International Nuclear Information System (INIS)

Fujiwara, Masami; Itoh, Satoshi; Matsuoka, Keisuke; Matsuura, Kiyokata; Miyamoto, Kenro.

1974-11-01

Conceptional design of a system for feedback control of the plasma position in a toroidal discharge is described. It is expected that a resistive shell and an external vertical field controlled by a system consisting of a digital computer and phase-controlled thyristors can suppress the plasma displacement down to 10% of that in the case where the external control system is not operated. (auth.)

Control rod driving mechanism of reactor, control device and operation method therefor

International Nuclear Information System (INIS)

Ariyoshi, Masahiko; Matsumoto, Fujio; Matsumoto, Koji; Kinugasa, Kunihiko; Nara, Yoshihiko; Otama, Kiyomaro; Mikami, Takao

1998-01-01

The present invention provides a device for and a method of directly driving control rods of an FBR type reactor linearly by a cylinder type linear motor while having a driving shaft as an electric conductor. Namely, a linear induction motor drives a driving shaft connected with a control rod and vertically moving the control rod by electromagnetic force as an electric conductor. The position of the control rod is detected by a position detector. The driving shaft is hung by a wire by way of an electromagnet which is attachably/detachably held. With such a constitution, the driving shaft connected with the control rod can be vertically moved linearly, stopped or kept. Since they can be driven smoothly at a wide range speed, the responsibility and reliability of the reactor operation can be improved. In addition, since responsibility of the control rod operation is high, scram can be conducted by the linear motor. Since the driving mechanism can be simplified, maintenance and inspection operation can be mitigated. (I.S.)

Design of a reactor inlet temperature controller for EBR-2 using state feedback

International Nuclear Information System (INIS)

Vilim, R.B.; Planchon, H.P.

1990-01-01

A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs

Overview of environmental control aspects for the gas-cooled fast reactor

International Nuclear Information System (INIS)

Nolan, A.M.

1981-05-01

Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

Reactor power control device in BWR power plant

International Nuclear Information System (INIS)

Kurosawa, Tsuneo.

1997-01-01

The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

International Nuclear Information System (INIS)

Eberle, C.S.; Dean, E.M.; Angelo, P.L.

1995-01-01

A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

Control system of power supply for resistance welding machine

Directory of Open Access Journals (Sweden)

Світлана Костянтинівна Поднебенна

2017-06-01

Full Text Available This article describes the existing methods of heat energy stabilizing, which are realized in thyristor power supplies for resistance welding machines. The advantages and features of thyristor power supplies have been described. A control system of power supply for resistance welding machine with stabilization of heat energy in a welding spot has been developed. Measurements are performed in primary winding of a welding transformer. Weld spot heating energy is calculated as the difference between the energy, consumed from the mains, and the energy losses in the primary and secondary circuits of the welding transformer as well as the energy losses in the transformer core. Algorithms of digital signal processing of the developed control system are described in the article. All measurements and calculations are preformed automatically in real-time. Input signals to the control system are: transformer primary voltage and current, temperature of the welding circuit. The designed control system ensures control of the welding heat energy and is not influenced by the supply voltage and impedance changes caused by insertion of the ferromagnetic mass in the welding circuit, the temperature change during the welding process. The developed control system for resistance welding machine makes it possible to improve the quality of welded joints, increase the efficiency of the resistance welding machine

On-line method to identify control rod drops in Pressurized Water Reactors

International Nuclear Information System (INIS)

Souza, T.J.; Martinez, A.S.; Medeiros, J.A.C.C.; Palma, D.A.P.; Gonçalves, A.C.

2014-01-01

Highlights: • On-line method to identify control rod drops in PWR reactors. • Identification method based on the readings of the ex-core detector. • Recognition of the patterns in the ex-core detector responses. - Abstract: A control rod drop event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimise undesirable effects in such a scenario. The goal of this work is to develop an online method to identify control rod drops in PWR reactors. The method entails the construction of a tool based on ex-core detector responses. It proposes to recognize patterns in the neutron ex-core detectors responses and thus to make an online identification of a control rod drop in the core during the reactor operation. The results of the study, as well as the behaviour of the detector responses demonstrated the feasibility of this method