Mechanical properties of concrete for power reactor at high temperatures

International Nuclear Information System (INIS)

Kawase, Kiyotaka; Tanaka, Hitoshi; Nakano, Masayuki

1985-01-01

The purpose of this study is to investigate the mechanical properties of concrete for power reactor at high temperature. This paper presents the creep behavior of concrete at high temperature and the cause by which a specified aggregate is broken at a specified high temperature. The creep coefficient at high temperature is smaller than that at ordinary temperature. (author)

Engineering and physics of high-power-density, compact, reversed-field-pinch fusion reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Krakowski, R.A.; Schultz, K.R.; Steiner, D.

1989-01-01

The technical feasibility and key developmental issues of compact, high-power-density Reversed-Field-Pinch (RFP) reactors are the primary results of the TITAN RFP reactor study. Two design approaches emerged, TITAN-I and TITAN-II, both of which are steady-state, DT-burning, circa 1000 MWe power reactors. The TITAN designs are physically compact and have a high neutron wall loading of 18 MW m 2 . Detailed analyses indicate that: a) each design is technically feasible; b) attractive features of compact RFP reactors can be realized without sacrificing the safety and environmental potential of fusion; and c) major features of this particular embodiment of the RFP reactor are retained in a design window of neutron wall loading ranging from 10 to 20 MW/m 2 . A major product of the TITAN study is the identification and quantification of major engineering and physics requirements for this class of RFP reactors. These findings are the focus of this paper. (author). 26 refs.; 4 figs.; 1 tab

Tendencies in operating power reactors

International Nuclear Information System (INIS)

Brinckmann, H.F.

1987-01-01

A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

Reactor power distribution monitor

International Nuclear Information System (INIS)

Hoizumi, Atsushi.

1986-01-01

Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

International Nuclear Information System (INIS)

Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

2010-01-01

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

Reactor G1: high power experiments

International Nuclear Information System (INIS)

Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

1957-01-01

The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

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.)

A compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

International Nuclear Information System (INIS)

Prelas, M.A.; Boody, F.P.; Zediker, M.S.

1987-01-01

An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

Pellet bed reactor for multi-modal space power

International Nuclear Information System (INIS)

Buden, D.; Williams, K.; Mast, P.; Mims, J.

1987-01-01

A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

A study of UO2 wafer fuel for very high-power research reactors

International Nuclear Information System (INIS)

Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

1983-01-01

The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

International Nuclear Information System (INIS)

Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

2000-01-01

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

Nuclear power reactors of new generation

International Nuclear Information System (INIS)

Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

1988-01-01

The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

The feature of high flux engineering test reactor and its role in nuclear power development

International Nuclear Information System (INIS)

Lu Guangquan

1987-01-01

The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

Evolution of on-power fuelling machines on Canadian natural uranium power reactors

International Nuclear Information System (INIS)

Isaac, P.

1984-10-01

The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Power supplyer for reactor coolant recycling pump

International Nuclear Information System (INIS)

Nara, Hiroshi; Okinaka, Yo.

1991-01-01

The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

International Nuclear Information System (INIS)

Foerster, S.; Hewing, G.

1977-01-01

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

Directory of Open Access Journals (Sweden)

Zhe Dong

2014-11-01

Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

High-Temperature Reactor For Power Generation and District Heating

International Nuclear Information System (INIS)

Herzberger, Karlheinz

1987-01-01

The multinational BBC Brown Brave Group, which has its head-quarters in Baden/Switzerland, was founded in 1891. Its German company is Brown, Brave and CIEs AGM, Mannheim. The field of operation covers wide areas of electrical engineering: These includes mainly the manufacture of installations and equipment for the generation, conversion, distribution and utilization of electric power, with special emphasis on the capital goods sector. BBC erects turnkey power plants and manufactures electrical equipment for industrial plants and urban transport and main line trains. Also of major importance are standard electrical products such as motors, switches, cables, semiconductor devices as well as measuring and control equipment. In the field of nuclear power BBC is engaged in particular in the development and construction of high-temperature reactors for the generation of electric power and process heat. The following presentation gives a short view on the milestones of the HTR development achieved in 1987

Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

2000-03-01

A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

Low power reactor for remote applications

International Nuclear Information System (INIS)

Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

1985-01-01

A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

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.)

Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2002-07-01

The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

Gaseous fuel reactors for power systems

Science.gov (United States)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

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.)

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

Low power unattended defense reactor

International Nuclear Information System (INIS)

Kirchner, W.L.; Meier, K.L.

1984-01-01

A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, ''walk-away safe'' design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (OandM) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

Low power unattended defense reactor

International Nuclear Information System (INIS)

Kirchner, W.L.; Meier, K.L.

1984-01-01

A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

New generation of reactors for space power

International Nuclear Information System (INIS)

Boudreau, J.E.; Buden, D.

1982-01-01

Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

Highly Radiative Plasmas for Local Transport Studies and Power and Particle Handling in Reactor Regimes

International Nuclear Information System (INIS)

Bell, M.G.; Bell, R.E.; Budny, R.; Bush, C.E.; Hill, K.W.

1998-01-01

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into the Tokamak Fusion Test Reactor (TFTR) supershots and high-l(subscript) plasmas. At neutral beam injection (NBI) powers P(subscript B) greater than or equal to 30 MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both deuterium (D) and deuterium-tritium (DT) plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in the International Thermonuclear Experimental Reactor (ITER). The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

An overview of future sustainable nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

2013-07-01

In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

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.)

Power reactors in member states

International Nuclear Information System (INIS)

1975-01-01

This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Reactor power reduction system and method

International Nuclear Information System (INIS)

Bruno, S.J.; Dunn, S.A.; Raber, M.

1978-01-01

A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

Considerations in the design of a high power medical isotope production reactor

International Nuclear Information System (INIS)

Ball, Russell M.; Nordyke, William H.; Brown, Roy

2002-01-01

For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

Nuclear reactor application for high temperature power industrial processes

International Nuclear Information System (INIS)

Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

1977-01-01

This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

Status report on nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-01-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

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.)

Modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shepherd, L.R.

1988-01-01

The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

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.)

MIT research reactor. Power uprate and utilization

Energy Technology Data Exchange (ETDEWEB)

Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

2012-03-15

The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

Method of estimating thermal power distribution of core of BWR type reactor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1982-01-01

Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

HTGR nuclear power plants: features of the VGR-50 high temperature reactor

International Nuclear Information System (INIS)

Glebov, V.P.; Bogoyavlenskii, R.G.; Glushkov, E.S.; Grebennik, V.N.; Ponomarev-Stepnoi, N.N.; Vinogradov, V.P.

1983-01-01

Current developmental trends in the power industry are guided to an appreciable extent by the increasing shortages of fossil fuels (coal, petroleum, natural gas) and by ecological problems. Assuming a continuing trend in worldwide consumption of energy resources, we see the electric power industry using up 20%, the other 80% (petroleum, coal, natural gas) going into generating industrial process heat and space heat, transportation, the chemical processing industry, the metallurgical industry, and other branches of industry. In the future, nuclear power will have the job of not only meeting the needs of the electric power industry, but also generating process heat. The most promising type of nuclear power plant available for solving complex problems in generation of electric power and heat for technological processes in the metallurgical processing industry and chemical processing industry is the one based around high-temperature reactors

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

A method of reactor power decrease by 2DOF control system during BWR power oscillation

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Suzuki, Katsuo

1998-09-01

Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Cermet-fueled reactors for multimegawatt space power applications

International Nuclear Information System (INIS)

Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

1988-01-01

The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

Sodium fast reactor power monitoring using gamma spectrometry

Energy Technology Data Exchange (ETDEWEB)

Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

2009-06-15

This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

Fusion reactor development using high power particle beams

International Nuclear Information System (INIS)

Ohara, Y.

1990-01-01

The present paper outlines major applications of the ion source/accelerator to fusion research and also addresses the present status and future plans for accelerator development. Applications of ion sources/accelerators for fusion research are discussed first, focusing on plasma heating, plasma current drive, plasma current profile control, and plasma diagnostics. The present status and future plan of ion sources/accelerators development are then described focusing on the features of existing and future tokamak equipment. Positive-ion-based NBI systems of 100 keV class have contributed to obtaining high temperature plasmas whose parameters are close to the fusion break-even condition. For the next tokamak fusion devices, a MeV class high power neutral beam injector, which will be used to obtain a steady state burning plasma, is considered to become the primary heating and current drive system. Development of such a system is a key to realize nuclear fusion reactor. It will be entirely indebted to the development of a MeV class high current negative deuterium ion source/accelerator. (N.K.)

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-01-01

CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

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

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

Science.gov (United States)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

Advanced in-core monitoring system for high-power reactors

International Nuclear Information System (INIS)

Mitin, V.I.; Alekseev, A.N.; Golovanov, M.N.; Zorin, A.V.; Kalinushkin, A.E.; Kovel, A.I.; Milto, N.V.; Musikhin, A.M.; Tikhonova, N.V.; Filatov, V.P.

2006-01-01

This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

International Nuclear Information System (INIS)

Carter, R.E.

1985-01-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

Energy Technology Data Exchange (ETDEWEB)

Carter, R E

1985-07-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

Reactor power distribution monitor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems

International Nuclear Information System (INIS)

Parlos, A.G.; Khan, E.U.; Frymire, R.; Negron, S.; Thomas, J.K.; Peddicord, K.L.

1991-01-01

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a neutronic design trade-off study using a two-dimensioinal neutron transport model, as well as a simplified one-dimensional thermal analysis of the reactor core. In arriving at the most desirable configuration, various options have been considered and analyzed, and their advantages/disadvantages have been compared. However, because of space limitation, only the most favorable reactor configuration is presented in this summary

High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

Science.gov (United States)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

International Nuclear Information System (INIS)

Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

1997-01-01

Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

Power reactors in Member States. 1978 edition

International Nuclear Information System (INIS)

1978-01-01

The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Reactor power monitoring device

International Nuclear Information System (INIS)

Kono, Shigehiro.

1990-01-01

Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

Small space reactor power systems for unmanned solar system exploration missions

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model

Gas-core reactor power transient analysis. Final report

International Nuclear Information System (INIS)

Kascak, A.F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

International Nuclear Information System (INIS)

1986-06-01

The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-06-15

The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

Comparison of Pickering NGS performance with world power reactors, 1977

International Nuclear Information System (INIS)

Buhay, S.

Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

Mathematical game type optimization of powerful fast reactors

International Nuclear Information System (INIS)

Pavelesku, M.; Dumitresku, Kh.; Adam, S.

1975-01-01

To obtain maximum speed of putting into operation fast breeders it is recommended on the initial stage of putting into operation these reactors to apply lower power which needs less fission materials. That is why there is an attempt to find a configuration of a high-power reactor providing maximum power for minimum mass of fission material. This problem has a structure of the mathematical game with two partners of non-zero-order total and is solved by means of specific aids of theory of games. Optimal distribution of fission and breeding materials in a multizone reactor first is determined by solution of competitive game and then, on its base, by solution of the cooperation game. The second problem the solution for which is searched is developed from remark on the fact that a reactor with minimum coefficient of flux heterogenity has a configuration different from the reactor with power coefficient heterogenity. Maximum burn-up of fuel needs minimum heterogenity of the flux coefficient and the highest power level needs minimum coefficient of power heterogenity. That is why it is possible to put a problem of finding of the reactor configuration having both coefficients with minimum value. This problem has a structure of a mathematical game with two partners of non-zero-order total and is solved analogously giving optimal distribution of fuel from the new point of view. In the report is shown that both these solutions are independent which is a result of the aim put in the problem of optimization. (author)

Study on the power control system for NPP power unit with the WWER-440 reactor

International Nuclear Information System (INIS)

Aleksandrova, N.D.; Naumov, A.V.

1981-01-01

Results of model investigations into basic version of the power control systems (PCS) conformably to the WWER-440 NPP power unit are stated. Transient processes in the power unit system when being two PCS versions during perturbations of different parameters: unit power, vapour pressure or position of control rods have been simulated. Investigations into the different PCS versions show that quality of operation of a traditional scheme with a turbine power controller and reactor pressure controller can be significantly improved with the introduction of a high-speed signal of pressure into the reactor controller. The PCS version with the compensation of interrelations between the turbine and reactor controllers constructed according to the same principles as the standard schemes of power units of thermal electric power plant is perspective as well [ru

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

Directory of Open Access Journals (Sweden)

Fic Adam

2015-03-01

Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)

Development and evaluation of a novel low power, high frequency piezoelectric-based ultrasonic reactor for intensifying the transesterification reaction

Directory of Open Access Journals (Sweden)

Mortaza Aghbashlo

2016-12-01

Full Text Available In this study, a novel low power, high frequency piezoelectric-based ultrasonic reactor was developed and evaluated for intensifying the transesterification process. The reactor was equipped with an automatic temperature control system, a heating element, a precise temperature sensor, and a piezoelectric-based ultrasonic module. The conversion efficiency and specific energy consumption of the reactor were examined under different operational conditions, i.e., reactor temperature (40‒60 °C, ultrasonication time (6‒10 min, and alcohol/oil molar ratio (4:1‒8:1. Transesterification of waste cooking oil (WCO was performed in the presence of a base-catalyst (potassium hydroxide using methanol. According to the obtained results, alcohol/oil molar ratio of 6:1, ultrasonication time of 10 min, and reactor temperature of 60 °C were found as the best operational conditions. Under these conditions, the reactor converted WCO to biodiesel with a conversion efficiency of 97.12%, meeting the ASTM standard satisfactorily, while the lowest specific energy consumption of 378 kJ/kg was also recorded. It should be noted that the highest conversion efficiency of 99.3 %, achieved at reactor temperature of 60 °C, ultrasonication time of 10 min, and alcohol/oil molar ratio of 8:1, was not favorable as the associated specific energy consumption was higher at 395 kJ/kg. Overall, the low power, high frequency piezoelectric-based ultrasonic module could be regarded as an efficient and reliable technology for intensifying the transesterification process in terms of energy consumption, conversion efficiency, and processing time, in comparison with high power, low frequency ultrasonic system reported previously. Finally, this technology could also be considered for designing, developing, and retrofitting chemical reactors being employed for non-biofuel applications as well.

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.)

Research and development into power reactor fuel performance

International Nuclear Information System (INIS)

Notley, M.J.F.

1983-07-01

The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

International Nuclear Information System (INIS)

Liu Yu; Dong Yujie

2011-01-01

In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, W.H.

1997-01-01

Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

Power conditioning system for a nuclear reactor

International Nuclear Information System (INIS)

Higashigawa, Yuichi; Joge, Toshio.

1981-01-01

Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

Impacts on power reactor health physics programs

International Nuclear Information System (INIS)

Meyer, B.A.

1991-01-01

The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

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

Thermionic reactors for space nuclear power

Science.gov (United States)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Selected power reactor projects in Canada and the United States of America

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-11-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

Selected power reactor projects in Canada and the United States of America

International Nuclear Information System (INIS)

1964-01-01

As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

International Nuclear Information System (INIS)

Luo, Chending; Zhao, Fuqiang; Zhang, Na

2014-01-01

Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

A high temperature reactor for ship propulsion

International Nuclear Information System (INIS)

Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

2002-01-01

The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

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.)

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

A Gas-Cooled Reactor Surface Power System

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

1998-11-09

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

TRIGA research reactors with higher power density

International Nuclear Information System (INIS)

Whittemore, W.L.

1994-01-01

The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (≅48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D 2 O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 10 14 n/cm 2 ·s and fast fluxes (>0.1 MeV) of 2 x 10 14 n/cm 2 ·s. The core centerline thermal neutron flux in the D 2 O reflector is about 2 x 10 14 n/cm 2 ·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D 2 O reflector of 3 x 10 14 n/cm 2 ·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

Power distribution forecasting device for reactors

International Nuclear Information System (INIS)

Tsukii, Makoto

1981-01-01

Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

Power distribution monitor in a nuclear reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1983-01-01

Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

Power reactors in Member States. 1979 edition

International Nuclear Information System (INIS)

1979-01-01

This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

High temperature reactor safety and environment

International Nuclear Information System (INIS)

Brisbois, J.; Charles, J.

1975-01-01

High-temperature reactors are endowed with favorable safety and environmental factors resulting from inherent design, main-component safety margins, and conventional safety systems. The combination of such characteristics, along with high yields, prove in addition, that such reactors are plagued with few problems, can be installed near users, and broaden the recourse to specific power, therefore fitting well within a natural environment [fr

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Power unit with GT-MHR reactor plant for electricity production and district heating

International Nuclear Information System (INIS)

Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

2000-01-01

Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

Investigation of applications for high-power, self-critical fissioning uranium plasma reactors. Final technical report

International Nuclear Information System (INIS)

Rodgers, R.J.; Latham, T.S.; Krascella, N.L.

1976-09-01

Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction. (Author)

Power reactor information system (PRIS)

International Nuclear Information System (INIS)

1989-06-01

Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

Sodium fast reactor power monitoring using {sup 20}F tagging agent

Energy Technology Data Exchange (ETDEWEB)

Coulon, R.; Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, Centre de Saclay, 91191 Gif sur Yvette Cedex (France); Ban, G. [ENSICAEN, F-14050 Caen (France); Dumarcher, V.; Brau, H. P.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A. M.; Montagu, T.; Dautremer, T.; Barat, E.

2009-07-01

This work deals with the use of gamma spectrometry to monitor the fourth generation sodium fast reactor (SFR) power. Simulation part has shown that power monitoring in short response time and with high accuracy is possible measuring delayed gamma emitters produced in the liquid sodium. An experimental test is under preparation at French SFR Phenix experimental reactor to validate simulation studies. Physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as sodium velocity, atomic densities, neutron spectra and incident neutron cross-sections of fission reactions, and also sodium activation reactions producing gamma emitters. Then, a thermal hydraulic transfer function was used for taking into account primary sodium flow in our calculations. Gamma spectra were then determined by Monte-Carlo simulations. The experiment will be set during the reactor 'ultimate testing'. The Delayed Neutron Detection (DND) system cell has been chosen as the best available primary sodium sample for gamma power monitoring on Phenix reactor due to short sodium transit time from reactor core to measurement sample and homogenized sampling in the reactor hot pool. The main gamma spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The signal is then processed by a digital signal processing system (called Adonis) which always gives optimum answer even for high count rate and various time activity measurements. For power monitoring problematic, use of a short decay period gamma emitter as the {sup 20}F will allow to obtain a very fast response system without cumulative and flow distortion effects. These works introduce advantages and performances of this new power monitoring system for future SFR. (authors)

Thermal Energetic Reactor with High Reproduction of Fission Materials

Directory of Open Access Journals (Sweden)

Vladimir M. Kotov

2012-01-01

On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

Modular high-temperature reactor launched (and wallchart)

International Nuclear Information System (INIS)

Steinwarz, W.

1987-01-01

In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Fuel element concept for long life high power nuclear reactors

Science.gov (United States)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Modular Lead-Bismuth Fast Reactors in Nuclear Power

Directory of Open Access Journals (Sweden)

Vladimir Petrochenko

2012-09-01

Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

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)

Power reactors operational diagnosis

International Nuclear Information System (INIS)

Dach, K.; Pecinka, L.

1976-01-01

The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

Small and medium power reactors 1987

International Nuclear Information System (INIS)

1987-12-01

This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

Reactor power system deployment and startup

International Nuclear Information System (INIS)

Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

Small and medium power reactors 1987

Science.gov (United States)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

State of development of high temperature gas-cooled reactors in foreign countries

International Nuclear Information System (INIS)

Sudo, Yukio

1990-01-01

Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

THE PHASE REACTOR INDUCTANCE SELECTION TECHNIQUE FOR POWER ACTIVE FILTER

Directory of Open Access Journals (Sweden)

D. V. Tugay

2016-12-01

Full Text Available Purpose. The goal is to develop technique of the phase inductance power reactors selection for parallel active filter based on the account both low-frequency and high-frequency components of the electromagnetic processes in a power circuit. Methodology. We have applied concepts of the electrical circuits theory, vector analysis, mathematical simulation in Matlab package. Results. We have developed a new technique of the phase reactors inductance selection for parallel power active filter. It allows us to obtain the smallest possible value of THD network current. Originality. We have increased accuracy of methods of the phase reactor inductance selection for power active filter. Practical value. The proposed technique can be used in the design and manufacture of the active power filter for real objects of energy supply.

The zero power reactor SUR and its application

International Nuclear Information System (INIS)

Wesser, U.

1986-01-01

This low-power reactor, rated nominally at 100 milliwatts, has a cylindrical core of 26 cm in diameter and 24 cm high consisting of U 3 O 8 powder in a polyethylene matrix. The fuel is 20 percent enriched and the critical mass about 700 g. The excess reactivity is about 3 mk. The reactivity is controlled by two cadmium sheets in addition to a back-up system that drops the inner reflector. The reactor has no active cooling system. Personnel costs include a supervisor and an operator. The reactor is used for training in Reactor Theory (including use of a neutron chopper), reactor kinetics, nuclear technology, reactor operations and for doctoral thesis research. (author)

The Swedish Zero Power Reactor R0

Energy Technology Data Exchange (ETDEWEB)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-15

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

The Swedish Zero Power Reactor R0

International Nuclear Information System (INIS)

Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

1961-05-01

The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

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)

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.)

Parameter study toward economical magnetic fusion power reactors

International Nuclear Information System (INIS)

Yoshida, Tomoaki; Okano, Kunihiko; Nanahara, Toshiya; Hatayama, Akiyoshi; Yamaji, Kenji; Takuma, Tadashi.

1996-01-01

Although the R and D of nuclear fusion reactors has made a steady progress as seen in ITER project, it has become of little doubt that fusion power reactors require hugeness and enormous amount of construction cost as well as surmounting the physics and engineering difficulties. Therefore, it is one of the essential issues to investigate the prospect of realizing fusion power reactors. In this report we investigated the effects of physics and engineering improvements on the economics of ITER-like steady state tokamak fusion reactors using our tokamak system and costing analysis code. With the results of this study, we considered what is the most significant factor for realizing economical competitive fusion reactors. The results show that with the conventional TF coil maximum field (12T), physics progress in β-value (or Troyon coefficient) has the most considerable effect on the reduction of fusion plant COE (Cost of Electricity) while the achievement of H factor = 2-3 and neutron wall load =∼5MW/m 2 is necessary. The results also show that with the improvement of TF coil maximum field, reactors with a high aspect ratio are economically advantageous because of low plasma current driving power while the improvement of current density in the conductors and yield strength of support structures is indispensable. (author)

SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

Energy Technology Data Exchange (ETDEWEB)

Williamson, R. A.

1963-10-15

The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

Spiegelberg, R.

1992-01-01

The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

Compact power reactor

International Nuclear Information System (INIS)

Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

Self-powered detectors for power reactors: an overview

International Nuclear Information System (INIS)

Ma, J.

2006-01-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Self-powered detectors for power reactors: an overview

Energy Technology Data Exchange (ETDEWEB)

Ma, J. [Univ. of Western Ontario, Dept. of Mechanical and Materials Engineering, London, Ontario (Canada)]. E-mail: jma64@uwo.ca

2006-07-01

In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

Fractals in Power Reactor Noise

International Nuclear Information System (INIS)

Aguilar Martinez, O.

1994-01-01

In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

Nuclear Power Reactors in the World. 2013 Ed

International Nuclear Information System (INIS)

2013-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-third edition of Reference Data Series No. 2 provides a detailed comparison of various statistics through 31 December 2012. The tables and figures contain the following information: - General statistics on nuclear reactors in IAEA Member States; - Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; - Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA's Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects data through designated national correspondents in Member States

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

Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors

International Nuclear Information System (INIS)

Gupta, S.K.; Kumar, Rajesh; Gaikwad, A.J.; Majumdar, P.; Agrawal, R.A.

2004-01-01

Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors (220 MWe) the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper. (author)

New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

International Nuclear Information System (INIS)

Dussol, R.J.

2003-01-01

In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

Basic policy of maintenance for the power conversion system of the gas turbine high temperature reactor 300 (GTHTR300)

International Nuclear Information System (INIS)

Kosugiyama, Shinichi; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, Xing; Katanishi, Shoji; Takada, Shoji

2003-01-01

Basic policy of maintenance was determined for major equipment in the power conversion system of the Gas Turbine High Temperature Reactor 300 (GTHTR300). It was developed based on the current maintenance practice in Light Water Reactors (LWRs), High Temperature Engineering Test Reactor (HTTR) and conventional combined cycle power plants while taking into account of unique design features of GTHTR300. First, potential degradation phenomena in operations were identified and corresponding maintenance approaches were proposed for the equipment. Such degradations encountered typically in LWRs as corrosion, erosion and stress corrosion cracking are unlikely to occur since the working fluid of GTHTR300 is inert helium. Main causes of the degradations are high operating temperature and pressure. The gas turbine, compressor, generator, control valves undergo opening and dismantling maintenance in a suitable time interval. The power conversion vessel, heat exchanger vessel, primary system piping and heat exchanging tubes of precooler are subjected to in-service inspections similar to those done in LWRs. As turbine blades represent the severest material degradation because of their high-temperature and high-stress operating conditions, a lifetime management scheme was suggested for them. The longest interval of open-casing maintenance of the gas turbine is estimated to be six to seven years from technical point of view. Present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

Directory of Open Access Journals (Sweden)

Zhe Dong

2014-02-01

Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

Reference reactor module for NASA's lunar surface fission power system

International Nuclear Information System (INIS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Thermal Energetic Reactor with High Reproduction of Fission Materials

International Nuclear Information System (INIS)

Kotov, V.M.

2012-01-01

Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, William H. Sr.

1998-01-01

Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

BN-1200 Reactor Power Unit Design Development

International Nuclear Information System (INIS)

Vasilyev, B.A.; Shepelev, S.F.; Ashirmetov, M.R.; Poplavsky, V.M.

2013-01-01

Main goals of BN-1200 design: • Develop a reliable new generation reactor plant for the commercial power unit with fast reactor to implement the first-priority objectives in changing over to closed nuclear fuel cycle; • Improve technical and economic indices of BN reactor power unit to the level of those of Russian VVER of equal power; • Enhance the safety up to the level of the requirements for the 4th generation RP

Tandem mirror reactor power balance studies

International Nuclear Information System (INIS)

Gorker, G.E.; Perkins, L.J.

1985-01-01

A tandem mirror reactor (TMR) power plant balance model has been developed and is now being used as a computer aid for performing parametric studies. End-cell power injection into the plasma and the physics thermal Q are used to determine the fusion power. About 80% of the fusion power is transferred by high-energy neutrons to the blanket modules and structures. The other 20% of the fusion power in the high-energy alpha particles is used to heat the deuterium-tritium (D-T) plasma. Most of the plasma-ionized particles transfer their energy to the halo dumps and direct converters. The plant efficiency is calculated for three different system cycles: (1) the pressurized water/saturated steam cycle; (2) the superheated steam cycle; and (3) the more complex superheat/reheat cycle. There is a signficiant improvement in plant efficiency as the electrical power multiplication factor and steam cycle efficiency increases

Tokamak experimental power reactor

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1978-01-01

A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

Medium-Power Lead-Alloy Reactors: Missions for This Reactor Technology

International Nuclear Information System (INIS)

Todreas, Neil E.; MacDonald, Philip E.; Hejzlar, Pavel; Buongiorno, Jacopo; Loewen, Eric P.

2004-01-01

A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [∼100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forced versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant.These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO 2 power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

Science.gov (United States)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

Nuclear Power Reactors in the World. 2014 Ed

International Nuclear Information System (INIS)

2014-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-fourth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2013. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects this data through designated national correspondents in Member States

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

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.)

How power is generated in a nuclear reactor

International Nuclear Information System (INIS)

Swaminathan, V.

1978-01-01

Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

Fuel element for high-temperature nuclear power reactors

International Nuclear Information System (INIS)

Schloesser, J.

1974-01-01

The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

Reaction-rate coefficients, high-energy ions slowing-down, and power balance in a tokamak fusion reactor plasma

International Nuclear Information System (INIS)

Tone, Tatsuzo

1978-07-01

Described are the reactivity coefficient of D-T fusion reaction, slowing-down processes of deuterons injected with high energy and 3.52 MeV alpha particles generated in D-T reaction, and the power balance in a Tokamak reactor plasma. Most of the results were obtained in the first preliminary design of JAERI Experimental Fusion Reactor (JXFR) driven with stationary neutral beam injection. A manual of numerical computation program ''BALTOK'' developed for the calculations is given in the appendix. (auth.)

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

International Nuclear Information System (INIS)

Popa-Simil, Liviu

2008-01-01

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

Advanced High Temperature Reactor Systems and Economic Analysis

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

2011-09-01

The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

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)

Multiple microprocessor based nuclear reactor power monitor

International Nuclear Information System (INIS)

Lewis, P.S.; Ethridge, C.D.

1979-01-01

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

The program of reactors and nuclear power plants

International Nuclear Information System (INIS)

Calabrese, Carlos R.

2001-01-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

Summary of space nuclear reactor power systems, 1983--1992

Energy Technology Data Exchange (ETDEWEB)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Summary of space nuclear reactor power systems, 1983--1992

International Nuclear Information System (INIS)

Buden, D.

1993-01-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power

Verification of the linearity of the IPR-R1 TRIGA reactor power channels

International Nuclear Information System (INIS)

Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

2013-01-01

The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

Reactors of different types in the world nuclear power

International Nuclear Information System (INIS)

Simonov, K.V.

1991-01-01

The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

Reactor Power Meter type SG-8

Energy Technology Data Exchange (ETDEWEB)

Glowacki, S W

1981-01-01

The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

Nuclear power reactors in the world. April 1990 ed.

International Nuclear Information System (INIS)

1990-01-01

This is the tenth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1989 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's power reactor information system (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States

Nuclear power reactors in the world. Apr 1991 ed.

International Nuclear Information System (INIS)

1991-01-01

This is the eleventh edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1990, on power reactors operating or under construction, and shut down; performance data on reactors operating in the Agency's Member States, as reported to the IAEA. This information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States. 5 figs, 19 tabs

Nuclear power reactors in the world. April 2005 ed

International Nuclear Information System (INIS)

2005-01-01

This is the twenty-fifth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: - General information as of the end of 2004 on power reactors operating or under construction, and shut down; - Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www.iaea.org/programmes/a2

Underground nuclear power station using self-regulating heat-pipe controlled reactors

International Nuclear Information System (INIS)

Hampel, V.E.

1989-01-01

The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

Underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Progress in space nuclear reactor power systems technology development - The SP-100 program

Science.gov (United States)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Reactor power peaking information display

International Nuclear Information System (INIS)

Book, T.L.; Kochendarfer, R.A.

1986-01-01

This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

International Nuclear Information System (INIS)

Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

2010-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

Evaluation for reasonableness of power conversion system concepts in the gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Minatsuki, I.; Mizokami, Y.

2007-01-01

The conceptual design study for the Gas Turbine High Temperature Reactor (GTHTR300) was completed in 2004. In GTHTR300, SECO (Simple, Economical Competitiveness and Originality) is advocated as design philosophy in order to minimize technical and economical requirement. Furthermore the design of the GTHTR300 was developed with reflecting various view points from utilities, manufacturers and research organizations. In GTHTR300, the horizontal turbo machine rotor, the turbo machine in a separated vessel, the turbo machine with single rotor, the generator inside the power conversion vessel, and the power conversion system without inter-coolers were selected as major power conversion system concepts. This paper describes the investigation and analysis about the major concepts of GTHTR300 power conversion system in order to evaluate reasonableness of GTHTR300 design approach and acceptability with using experience and engineering knowledge of Mitsubishi Heavy Industries, Ltd., which were accumulated through the activities of HTGR-GT and HTTR (High Temperature Engineering Test Reactor) designing, manufacturing, fabricating and testing. From the result of the evaluation, it was concluded that the selection of each concept in GTHTR300 was reasonable as based on the original design philosophy SECO. As a conclusion, we expect the GTHTR300 to become one of the most promising concepts for commercialization in near future. (authors)

Commissioning of the THTR-300-MWe prototype power plant - A milestone for further application of this high-temperature reactor line

International Nuclear Information System (INIS)

Simon, M.; Baust, E.; Schoening, J.

1986-10-01

With the completion of the THTR 300 and the development of the follow-on plant HTR 500, the BBC/HRB company group has taken the pebble bed high-temperature reactor to the threshold of the commercial stage. The HTR is an important innovation in the field of reactor technology which can play an important role in the intermediate and long-term supply of safe, environmental friendly and economic energy. The power level of 550 MW meets the requirements of the present energy market which shows a trend towards smaller power units as a result of grid size, investment effort, and the slower increase in electricity demand in industrial nations. The advantages of the high-temperature reactor, such as high thermal efficiency, low waste heat, low radiation exposure of operating and maintenance personnel, high inherent safety, simple mode of operation, flexible fuel cycle with the potential to extend fuel resources, high availability, are currently uncontested and will represent the future standards for the peaceful uses of nuclear energy. For special applications in industry (steam and electric power as a cogeneration product) and in case of special siting conditions (near industrial centers), BBC/HRB developed a small 100 MW HTR, which can also be constructed as a 200 MW twin plant at favorable cost conditions. For an economic use of domestic coal in a processed form, the HTR represents the optimum solution as to economic and environmental aspects as well as extension of resources, especially if combined with conventional gasification procedures and in direct application of nuclear process heat at high gas temperatures of about 950 deg. C. In this field the development of the heat-exchanging components remains to be completed, before commercial application will be possible. The HTR is particularly well suited for erection in developing countries and industrial threshold countries which turn to nuclear energy for the first time. On an international level the interest in the

Calibration of RB reactor power; Kalibrisanje snage reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Markovic, H; Ninkovic, M; Strugar, P; Dimitrijevic, Z; Takac, S; Stefanovic, D; Kocic, A; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1976-09-15

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8{radical}2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation.

An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

International Nuclear Information System (INIS)

Wang, L.C.

1984-01-01

The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

Science.gov (United States)

Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

1991-01-01

This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

Cummings, G.E.

1987-01-01

Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety

High-temperature reactor developments in the Netherlands

International Nuclear Information System (INIS)

Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van.

1996-01-01

The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.)

Critical Power Response to Power Oscillations in Boiling Water Reactors

International Nuclear Information System (INIS)

Farawila, Yousef M.; Pruitt, Douglas W.

2003-01-01

The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

International Nuclear Information System (INIS)

Minatsuki, Isao; Mizokami, Yorikata

2007-01-01

The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Output feedback dissipation control for the power-level of modular high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Dong, Z.

2011-01-01

Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR) is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis. (author)

Power Nuclear Reactors: technology and innovation for development in future

International Nuclear Information System (INIS)

Suarez Antola, R.

2009-01-01

The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

A computational modeling approach of the jet-like acoustic streaming and heat generation induced by low frequency high power ultrasonic horn reactors.

Science.gov (United States)

Trujillo, Francisco Javier; Knoerzer, Kai

2011-11-01

High power ultrasound reactors have gained a lot of interest in the food industry given the effects that can arise from ultrasonic-induced cavitation in liquid foods. However, most of the new food processing developments have been based on empirical approaches. Thus, there is a need for mathematical models which help to understand, optimize, and scale up ultrasonic reactors. In this work, a computational fluid dynamics (CFD) model was developed to predict the acoustic streaming and induced heat generated by an ultrasonic horn reactor. In the model it is assumed that the horn tip is a fluid inlet, where a turbulent jet flow is injected into the vessel. The hydrodynamic momentum rate of the incoming jet is assumed to be equal to the total acoustic momentum rate emitted by the acoustic power source. CFD velocity predictions show excellent agreement with the experimental data for power densities higher than W(0)/V ≥ 25kWm(-3). This model successfully describes hydrodynamic fields (streaming) generated by low-frequency-high-power ultrasound. Crown Copyright © 2011. Published by Elsevier B.V. All rights reserved.

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

WITAMIR-I: A tandem mirror power reactor

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Beyer, J.B.

1983-01-01

A conceptual design of a near term commercial tandem mirror power reactor will be presented. The basic configuration utilizes Yin-yang minimum B end plugs with inboard thermal barriers, which are pumped by neutral beam injection. The maximum magnetic fields are 6.1 T, 8.1 T and 15 T in the central cell, Yin-yang, and thermal barrier magnets, respectively. The blanket utilizes Pb 83 Li 17 as the coolant and breeder, and HT-9 as the structural material. This configuration yields a high energy multiplication (1.37), a sufficient tritium breeding ratio (1.07) and has a major advantage with respect to maintenance. A single stage direct convertor is used at one end and an electron thermal dump at the other end. The plasma Q is 28 at a fusion power level of 3000 MWsub(th); the net electrical output is 1530 MWe and the overall efficiency is 39%. Cost estimates indicate that WITAMIR-I is competitive with recent tokamak power reactor designs. (author)

Power reactor design trends

International Nuclear Information System (INIS)

Hogan, W.J.

1985-01-01

Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

Self-operation type power control device for nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Kanbe, Mitsuru.

1993-07-23

The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

Self-operation type power control device for nuclear reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru.

1993-01-01

The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

International Nuclear Information System (INIS)

Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

2007-01-01

The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold (1) efficient low cost energy generation and (2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

Transmutation of actinides in power reactors.

Science.gov (United States)

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

Use of sup(233)U for high flux reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Liem, P.H.

1991-01-01

The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

A series of lectures on operational physics of power reactors

International Nuclear Information System (INIS)

Mohanakrishnan, P.; Rastogi, B.P.

1982-01-01

This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

Nuclear Power Reactors in the World. 2016 Ed

International Nuclear Information System (INIS)

2016-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. This thirty-sixth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2015. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. Data is collected by the IAEA via designated national correspondents in Member States

Highly radiative plasmas for local transport studies and power and particle handling in reactor regimes

International Nuclear Information System (INIS)

Hill, K.W.; Bell, M.G.; Budny, R.

1999-01-01

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high-l i plasmas. At neutral beam injection (NBI) powers P B ≥ 30 MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. (author)

Highly radiative plasmas for local transport studies and power and particle handling in reactor regimes

International Nuclear Information System (INIS)

Hill, K.W.; Bell, M.G.; Budny, R.

2001-01-01

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high-l i plasmas. At neutral beam injection (NBI) powers P B ≤30MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. (author)

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Science.gov (United States)

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power...

Nuclear detectors for in-core power-reactors

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1979-12-01

Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

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.)

Design of megawatt power level heat pipe reactors

Energy Technology Data Exchange (ETDEWEB)

Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-12

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

Nuclear power for coexistence with nature, high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Tight aspect ratio tokamak power reactor with superconducting TF coils

International Nuclear Information System (INIS)

Nishio, S.; Tobita, K.; Konishi, S.; Ando, T.; Hiroki, S.; Kuroda, T.; Yamauchi, M.; Azumi, M.; Nagata, M.

2003-01-01

Tight aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were discarded. The key point was how to find the engineering design solution of the TF coil system with the high field and high current density. The coil system with the center post radius of less than 1 m can generate the maximum field of ∼ 20 T. This coil system causes a compact reactor concept, where the plasma major and minor radii of 3.75 m and 1.9 m, respectively and the fusion power of 1.8 GW. (author)

Nuclear power reactors in the world. Apr 1985 ed.

International Nuclear Information System (INIS)

1985-01-01

This is the fifth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which replaces the Agency's publication Power Reactors in Member States. This bulletin contains the following summarized information on nuclear power reactors in the world: General information as of the end of 1984 on reactors operating or under construction and such additional information on planned and shutdown reactors as is available; Performance data on major reactor types operating in the Agency's Member States. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of and operating experience with reactors

An inherently safe power reactor module

International Nuclear Information System (INIS)

Salerno, L.N.

1985-01-01

General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

Power reactor core safety research

International Nuclear Information System (INIS)

Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

1981-01-01

As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

R&D on high-power dc reactor prototype for ITER poloidal field converter

Energy Technology Data Exchange (ETDEWEB)

Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

2015-10-15

Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

Science.gov (United States)

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

The comparative analysis of safety and economic competitiveness of the advanced high-power reactor projects of NPP

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, U.M.; Philimonova, R.A.; Kichutkina, E.G.

2002-01-01

The comparative analysis results of the safety and economic competitiveness of the seven advanced large sized reactors projects (900 MW and more) are submitted in that report: EPR, Frameatome France, Siemens Germany; EP-1000 Westinghouse, USA and Genesi Italy; Candu 9, Atomic Energy of Canada Ltd; System 80 +, ABB, USA; KNGR, group NSSS Engineering and Development, Korea Power Engineering Company, Inc; APWR, Electric Power company, Japan Atomic Power Company, Mitsubishi Heavy Industries, Westinghouse Electric; and WWER-1000 (V-392), Atomenergoproject/Gidropress Russian Federation. According to the economic competitiveness of listed compared power reactors the 14 criteria of safety have been accepted. These criteria: 1. Features of the barrier system of 'defence-in-depth'. 2. The self-security of a reactor under increase of power and reactivity of a reactor, decrease of the expense and phase transformations of the reactor core coolant (presence of negative feedbacks). 3. Presence of the reactor shutdown systems responding principles of a variety, independence and reservation. Presence of the passive means of initiation and operation of the emergency protection. 4. The emergency cooling of the core of reactor. A presence of the passive means of cooling. Presence of the water reservation for the water supply of the different safety systems. 5. The emergency electrical supply, its reliability and degree of reservation. 6. The prevention measures of the heavy accident with the melt core. The decrease of the heavy accident probability. 7. The account of the heavy accident under development of the levels of protection. 8. The protection levels of NPP, the technological criteria of efficiency of the each safety barriers and the limiting radiation criteria for the each level of protection , in particular for the design-basis and beyond-design-basis accidents. 9. The measures for reduction of the heavy accident consequences. The management by the beyond

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

Safety Analysis Of Actinide Recycled Fast Power Reactor

International Nuclear Information System (INIS)

Taufik, Mohammad

2001-01-01

Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

Conversion of a reactor to partial power output

International Nuclear Information System (INIS)

Iljunin, W.G.; Kusnezow, I.A.; Murogow, W.M.; Schmelew, A.N.

1975-01-01

The method, among other things, involves an increase in the rate of secondary fissile material production in a fast breeder reactor if the flow of the working fluid through a turbine is reduced as a function of a given amount of reduction of the electric load. This objective will be served by a circuit and circuit variants, respectively, which include a high temperature cooling circuit with, for instance, a sodium cooled HTR, a low temperature cooling circuit with, for instance, a fast or thermal breeder reactor, a working fluid circuit with the turbine, and a heat consumption circuit. In the scheme suggested for operation in the partial power production mode it is envisaged that, as the electric load of the plant decreases the flow of the working fluid upstream of the turbine is kept constant by means of a control system in the working fluid circuit. Additional control systems are used to reduce the amount of heat transmitted by the breeder reactor to the working fluid. The excess amount of heat is distributed to the load connected. This again reduces the temperatures at the inlet and the outlet of the breeder reactor, thus raising its thermal power output. However, the flow through the breeder reactor remains constant all the time. (DG/RF) [de

Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

International Nuclear Information System (INIS)

Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

2008-01-01

The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

The system of the measurement of reactor power and the monitoring of core power distribution

International Nuclear Information System (INIS)

Li Xianfeng

1999-01-01

The author mainly describes the measurement of the reactor power and the monitoring of the core power distribution in DAYA BAY nuclear power plant, introduces the calibration for the measurement system. Ex-core nuclear instrumentation system (RPN) and LOCA surveillance system (LSS) are the most important system for the object. they perform the measurement of the reactor power and the monitoring of the core power distribution on-line and timely. They also play the important roles in the reactor control and the reactor protection. For the same purpose there are test instrumentation system (KME) and in-core instrumentation system (RIC). All of them work together ensuring the exact measurement and effective monitoring, ensuring the safety of the reactor power plant

Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

International Nuclear Information System (INIS)

Novelli, A.

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

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

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

The assessment of technological and safety aspects of small power reactor SMART

International Nuclear Information System (INIS)

Antariksawan, A.R.; Ekariansyah, Andi S.; Sony, D.T.; Suharno; Hastowo, Hudi

2002-01-01

This paper describes and discusses the technology and safety of small nuclear power plant SMART. The reactor SMART produces 300 MWth of power is cooled and moderated with light water and integral PWR type developed by KAERI. At present, the development activities had reached the end of basic design stage. The concept design of reactor SMART is based on safety enhancement, economic competitiveness and high performance. The fuel is uranium oxide with approximately 5% w/o enrichment. The safety characteristics of the core are shown with low power density around 62.6 W/cc, high negative reactivity coefficient, and high shutdown and thermal margin. Besides the inherent safety characteristics, SMART is equipped with engineered safety features and severe accident management system which are in compliance with the IAEA recommendations. The application of SMART for dual-purpose produces 90 Mwe and 40,000 to fresh water a day. Based on the technology and core characteristics of the reactor SMART, it is very interesting to be deeply assessed

Study of advanced fission power reactor development for the United States. Volume I

International Nuclear Information System (INIS)

1976-01-01

This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

An underground nuclear power station using self-regulating heat-pipe controlled reactors

Science.gov (United States)

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Investigations to the potential of the high temperature reactor for steam power processes with highest steam conditions and comparison with according conventional power plants

International Nuclear Information System (INIS)

Mondry, M.

1988-04-01

Already in the fifties conventional power plants with high parameters of the live steam were built to improve the total efficiency. The power plant with the highest steam conditions in the Federal Republic of Germany has 300 bar pressure and 600deg C temperature. Because of high material costs and other problems power plants with such high conditions were not continued to be built. Standard conditions of today's power plants are in the order of 180-250 bar pressure and 535deg C temperature. As the high temperature reactor is partly built up in another way than a conventional power plant, the results regarding the high steam parameters are not transferable. Possibilities for the technical realization of determined HTR-specific components are introduced and discussed. Then different HTR-power plants with steam conditions up to 350 bar pressure and 650deg C temperature are projected. Economical considerations show that an HTR with higher steam parameters brings financial profits. Further efficiency increase, which is possible by the high steam conditions, is shortly presented. The work ends with a technical and economical comparison of corresponding conventional power plants. (orig./UA) [de

Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

International Nuclear Information System (INIS)

Kusserow, J.; Gurr, W.; Pflug, H.

1985-05-01

Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

Nuclear-Mechanical Coupling: Small Amplitude Mechanical Vibrations and High Amplitude Power Oscillations in Nuclear Reactors

International Nuclear Information System (INIS)

Suarez Antola, R.

2008-11-01

The cores of nuclear reactors, including its structural parts and cooling fluids, are complex mechanical systems able to vibrate in a set of normal modes and frequencies, if suitable perturbed. The cyclic variations in the strain state of the core materials may produce changes in density. Changes in density modify the reactivity. Changes in reactivity modify thermal power. Modifications in thermal power produce variations in temperature fields. Variations in temperature produce variations in strain due to thermal-elastic effects. If the variation of the temperature field is fast enough and if the Doppler Effect and other stabilizing prompt effects in the fuel are weak enough, a fast oscillatory instability could be produced, coupled with mechanical vibrations of small amplitude. A recently constructed, simple mathematical model of nuclear reactor kinetics, that improves the one due to A.S. Thompson, is reviewed. It was constructed in order to study, in a first approximation, the stability of the reactor: a nonlinear nuclear-thermal oscillator (that corresponds to reactor point kinetics with thermal-elastic feedback and with frozen delayed neutron effects) is coupled nonlinearly with a linear mechanical-thermal oscillator (that corresponds to the first normal mode of mechanical vibrations excited by thermo-elastic effects). This mathematical model is studied here from the standpoint of mechanical vibrations. It is shown how, under certain conditions, a suitable mechanical perturbation could elicit fast and growing oscillatory instabilities in the reactor power. Applying the asymptotic method due to Krylov, Bogoliubov and Mitropolsky, analytical formulae that may be used in the calculation of the time varying amplitude and phase of the mechanical oscillations are given, as functions of the mechanical, thermal and nuclear parameters of the reactor. The consequences for the mechanical integrity of the reactor are assessed. Some conditions, mainly, but not exclusively

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1990-01-01

Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

Directory of Open Access Journals (Sweden)

Zhe Dong

2011-11-01

Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

Conceptual Study for development of a low power research reactor

International Nuclear Information System (INIS)

Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C.

2013-01-01

Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor

Gas core reactor power plants designed for low proliferation potential

International Nuclear Information System (INIS)

Lowry, L.L.

1977-09-01

The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

Reactor power region measuring device

International Nuclear Information System (INIS)

Kashiwa, Takao.

1996-01-01

The device of the present invention can rapidly detect abnormality of a local power region monitor (LPRM) even at a low power region caused such as upon start-up of a BWR type reactor. Namely, the present invention comprises (1) an LPRM detector for measuring neutron fluxes in the reactor, (2) a gamma thermo detector for calibrating the sensitivity of the LPRM detector, (3) a comparison circuit for comparing the detected values of the detectors (1) and (2), and (4) an alarm circuit for outputting an alarm when the comparative difference of the output of the circuit (3) exceeds a predetermined value. Signals of an alarm for a lower limit of the LPRM detector have been issued continuously upon start-up and shut down of the reactor since neutron fluxes in the reactor are reduced. However, the gamma thermo detector is always secured in the inside of the reactor different from a travelling-type incore probe monitor (TIP) disposed so far for the same purpose. Accordingly, the alarm generated upon usual start-up can be eliminated by comparing the detected values of the detector (2) and abnormality of the detector (1) can be rapidly detected by judging the abnormality of the comparative difference. (I.S.)

Nuclear safety. Concerns about the nuclear power reactors in Cuba

International Nuclear Information System (INIS)

Wells, Jim; Aloise, Gene; Flaherty, Thomas J.; Fitzgerald, Duane; Zavala, Mario; Hayward, Mary Alice

1992-09-01

In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

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.)

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.)

Reactor power distribution pattern judging device

International Nuclear Information System (INIS)

Ikehara, Tadashi.

1992-01-01

The judging device of the present invention comprises a power distribution readout system for intaking a power value from a fuel segment, a neural network having an experience learning function for receiving a power distribution value as an input variant, mapping it into a desirable property and self-organizing the map, and a learning date base storing a plurality of learnt samples. The read power distribution is classified depending on the similarity thereof with any one of representative learnt power distribution, and the corresponding state of the reactor core is outputted as a result of the judgement. When an error is found in the classified judging operation, erroneous cases are additionally learnt by using the experience and learning function, thereby improving the accuracy of the reactor core characteristic estimation operation. Since the device is mainly based on the neural network having a self-learning function and a pattern classification and judging function, a judging device having a human's intuitive pattern recognition performance and a pattern experience and learning performance is obtainable, thereby enabling to judge the state of the reactor core accurately. (N.H.)

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Directory of Open Access Journals (Sweden)

Ternovykh Mikhail

2017-01-01

Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Science.gov (United States)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Design study of ship based nuclear power reactor

International Nuclear Information System (INIS)

Su'ud, Zaki; Fitriyani, Dian

2002-01-01

Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

Reference reactor module for NASA's lunar surface fission power system

Energy Technology Data Exchange (ETDEWEB)

Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Instrumentation and control for reactor power setback in PFBR

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

2013-01-01

In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

Mesquita, Henrique F.A.; Ferreira, Andrea V.

2015-01-01

The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

Zirconium-hydride solid zero power reactor and its application research

International Nuclear Information System (INIS)

Lin Shenghuo; Luo Zhanglin; Su Zhuting

1994-10-01

The Zirconium Hydride Solid Zero Power Reactor built at China Institute of Atomic Energy is introduced. In the reactor Zirconium-hydride is used as moderator, plexiglass as reflector and U 3 O 8 with enrichment of 20% as the fuel, Since its initial criticality, the physical characteristics and safety features have been measured with the result showing that the reactor has sound stability and high sensitivity, etc. It has been successfully used for the personnel training and for the testing of reactor control instruments and experiment devices. It also presents the special advantage for the pre-research of some applications

Selection of power plant elements for future reactor space electric power systems

International Nuclear Information System (INIS)

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

International Nuclear Information System (INIS)

Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

1981-10-01

Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

Fractional power operation of tokamak reactors

International Nuclear Information System (INIS)

Mau, T.K.; Vold, E.L.; Conn, R.W.

1986-01-01

Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

High-intensity power-resolved radiation imaging of an operational nuclear reactor

OpenAIRE

Beaumont, Jonathan; Villa, Mario; Mellor, Matthew; Joyce, Malcolm John

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has bee...

Power limit and quality limit of natural circulation reactor

International Nuclear Information System (INIS)

Zhao Guochang; Ma Changwen

1997-01-01

The circulation characteristics of natural circulation reactor in boiling regime are researched. It is found that, the circulation mass flow rate and the power have a peak value at a mass quality respectively. Therefore, the natural circulation reactor has a power limit under certain technological condition. It can not be increased steadily by continually increasing the mass quality. Corresponding to this, the mass quality of natural circulation reactor has a reasonable limit. The relations between the maximum power and the reactor parameters, such as the resistance coefficient, the working pressure and so on, are analyzed. It is pointed out that the power limit of natural circulation reactor is about 1000 MW at present technological condition. Taking the above result and low quality stability experimental result into account, the authors recommend that the reasonable mass quality of natural circulation reactor working in boiling regime is from 2% to 3% under the researched working pressure

Statistic method of research reactors maximum permissible power calculation

International Nuclear Information System (INIS)

Grosheva, N.A.; Kirsanov, G.A.; Konoplev, K.A.; Chmshkyan, D.V.

1998-01-01

The technique for calculating maximum permissible power of a research reactor at which the probability of the thermal-process accident does not exceed the specified value, is presented. The statistical method is used for the calculations. It is regarded that the determining function related to the reactor safety is the known function of the reactor power and many statistically independent values which list includes the reactor process parameters, geometrical characteristics of the reactor core and fuel elements, as well as random factors connected with the reactor specific features. Heat flux density or temperature is taken as a limiting factor. The program realization of the method discussed is briefly described. The results of calculating the PIK reactor margin coefficients for different probabilities of the thermal-process accident are considered as an example. It is shown that the probability of an accident with fuel element melting in hot zone is lower than 10 -8 1 per year for the reactor rated power [ru

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

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

Control room conceptual design of nuclear power plant with multiple modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Jia Qianqian; Qu Ronghong; Zhang Liangju

2014-01-01

A conceptual design of the control room layout for the nuclear power plant with multiple modular high temperature gas-cooled reactors has been developed. The modular high temperature gas-cooled reactors may need to be grouped to produce as much energy as a utility demands to realize the economic efficiency. There are many differences between the multi-modular plant and the current NPPs in the control room. These differences may include the staffing level, the human-machine interface design, the operation mode, etc. The potential challenges of the human factor engineering (HFE) in the control room of the multi-modular plant are analyzed, including the operation workload of the multi-modular tasks, how to help the crew to keep situation awareness of all modules, and how to support team work, the control of shared system between modules, etc. A concept design of control room for the multi-modular plant is presented based on the design aspect of HTR-PM (High temperature gas-cooled reactor pebble bed module). HFE issues are considered in the conceptual design of control room for the multi-modular plant and some design strategies are presented. As a novel conceptual design, verifications and validations are needed, and focus of further work is sketch out. (author)

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

UF6 breeder reactor power plants for electric power generation

International Nuclear Information System (INIS)

Rust, J.H.; Clement, J.D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

Small reactor power systems for manned planetary surface bases

Energy Technology Data Exchange (ETDEWEB)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Small reactor power systems for manned planetary surface bases

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

Preliminary nuclear power reactor technology qualitative assessment for Malaysia

International Nuclear Information System (INIS)

Shamsul Amri Sulaiman

2011-01-01

Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

High power CW linac in PNC

International Nuclear Information System (INIS)

Toyama, S.; Wang, Y.L.; Emoto, T.

1994-01-01

Power Reactor and Nuclear Fuel Development Corporation (PNC) is developing a high power electron linac for various applications. The electron beam is accelerated in CW operation to get maximum beam current of 100 mA and energy of 10 MeV. Crucial components such as a high power L-band klystron and a high power traveling wave resonant ring (TWRR) accelerator guides were designed and manufactured and their performance were examined. These design and results from the recent high power RF tests were described in this paper. (author)

Decommissioning experience of the Japan power demonstration reactor

International Nuclear Information System (INIS)

Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

1992-01-01

Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

Rotating-bed reactor as a power source for EM gun applications

Energy Technology Data Exchange (ETDEWEB)

Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

1980-01-01

Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

Rotating-bed reactor as a power source for EM gun applications

International Nuclear Information System (INIS)

Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

1980-01-01

Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m 3 ), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300 0 K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications

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

Calculation device for fuel power history in BWR type reactors

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1980-01-01

Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

Simulation of decreasing reactor power level with BWR simulator

International Nuclear Information System (INIS)

Suwoto; Zuhair; Rivai, Abu Khalid

2002-01-01

Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

Development of methods for monitoring and controlling power in nuclear reactors

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole dos; Silva, Vitor Vasconcelos Araujo

2012-01-01

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235 U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of

Inherent Safety Feature of Hybrid Low Power Research Reactor during Reactivity Induced Accident

Energy Technology Data Exchange (ETDEWEB)

Kim, DongHyun; Yum, Soo Been; Hong, Sung Teak; Lim, In-Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Hybrid low power research reactor(H-LPRR) is the new design concept of low power research reactor for critical facility as well as education and training. In the case of typical low power research reactor, the purposes of utilization are the experiments for education of nuclear engineering students, Neutron Activation Analysis(NAA) and radio-isotope production for research purpose. H-LPRR is a light-water cooled and moderated research reactor that uses rod-type LEU UO{sub 2} fuels same as those for commercial power plants. The maximum core thermal power is 70kW and, the core is placed in the bottom of open pool. There are 1 control rod and 2 shutdown rods in the core. It is designed to cool the core by natural convection, retain negative feedback coefficient for entire fuel periods and operate for 20 years without refueling. Inherent safety in H-LPRR is achieved by passive design features such as negative temperature feedback coefficient and core cooling by natural convection during normal and emergency conditions. The purpose of this study is to find out that the inherent safety characteristics of H-LPRR is able to control the power and protect the reactor from the RIA(Reactivity induced accident). RIA analysis was performed to investigate the inherent safety feature of H-LPRR. As a result, it was found that the reactor controls its power without fuel damage in the event and that the reactor remains safe states inherently. Therefore, it is believed that high degree of safety inheres in H-LPRR.

Management of radioactive wastes at power reactor sites in India

International Nuclear Information System (INIS)

Amalraj, R.V.; Balu, K.

Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

Axial power monitoring uncertainty in the Savannah River Reactors

International Nuclear Information System (INIS)

Losey, D.C.; Revolinski, S.M.

1990-01-01

The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

Thermofluid-neutronic stability of the rotating, fluidized bed, space-power reactor

International Nuclear Information System (INIS)

Lee, C.C.; Jones, O.C.; Becker, M.

1993-01-01

A rotating fluidized bed nuclear reactor has the potential of being a vary attractive option for ultra-high power space systems, especially for propulsion. Research has already examined fuel bed expansion due to variations in state variables, propellant flow rate, and rotational speed, and has also considered problems related to thermal stress. This paper describes the results of a coupled thermofluid-neutronic analysis where perturbations in fuel bed height caused by maneuvering changes in operating conditions alter power levels due to varying absorption of neutrons which would otherwise leak from the system, mainly through the nozzle. This first analysis was not a detailed stability analysis. Rather, it utilized simplified neutronic methods, and was intended to provide an order-of-magnitude assessment of the stability of the reactor with the intention to determine whether or not stability might be a 'concept killer'. Stability was compared with a fixed-fuel-bed reactor of identical geometry for three different cases comprising a set of small, medium and large sizes/powers from 250 MW to 5 GW. It was found that power fluctuations in the fluidized bed reactor were larger by 100 db or more than expected in a packed bed reactor of the same geometry, but never resulted in power excursions. Margins to unit gain in some cases, however, were sufficiently small that the approximations in this quasi-2-dimensional model may not be sufficiently accurate to preclude significant excursions. (orig.)

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

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

Nuclear reactor capable of electric power generation during in-service inspection

International Nuclear Information System (INIS)

Nakamura, Shinsuke; Nogami, Hitoshi.

1992-01-01

The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

Hydrogen production from high temperature electrolysis and fusion reactor

International Nuclear Information System (INIS)

Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

1978-01-01

Production of hydrogen from high temperature electrolysis of steam coupled with a fusion reactor is studied. The process includes three major components: the fusion reactor, the high temperature electrolyzer and the power conversion cycle each of which is discussed in the paper. Detailed process design and analysis of the system is examined. A parametric study on the effect of process efficiency is presented

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.)

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

A comparative analysis of the domestic and foreign licensing processes for power and non-power reactors

International Nuclear Information System (INIS)

Joe, J. C.; Youn, Y. K.; Kim, W. S.; Kim, H. J.

2003-01-01

The System-integrated Modular Advanced Reactor (SMART), a small to medium sized integral type Pressurized Water Reactor (PWR) has been developed in Korea. Now, SMART-P, a 1/5 scaled-down of the SMART, is being developed for the purpose of demonstrating the safety and performance of SMART design. The SMART-P is a first-of-a-kind reactor which is utilized for the research and development of a power reactor. Since the licensing process of such a reactor is not clearly specified in the current Atomic Energy Act, a comparative survey and analysis of domestic and foreign licensing processes for power and non-power reactors has been carried out to develop the rationale and technical basis for establishing the licensing process of such a reactor. The domestic and foreign licensing processes of power and non-power reactors have been surveyed and compared, including those of the U.S.A., Japan, France, U.K., Canada, and IAEA. The general trends in nuclear reactor classification, licensing procedures, regulatory technical requirements, and other licensing requirements and regulations have been investigated. The results of this study will be used as the rationale and technical basis for establishing the licensing process of reactors at development stage such as SMART-P

Review of IVR-ERVC and using flooding concept for application to high power reactor

International Nuclear Information System (INIS)

Lee, Min ho; Heo, Hyo; Bang, In Cheol

2016-01-01

Accident scope will be limited in the RPV. For example, in case of Fukushima, they have difficulties for cleanup the accident and even catching the location of the melt-through corium. Therefore, IVR-ERVC is the right strategy for mitigation of the severe accident. However, in case of high power reactors, there is a Critical Heat Flux (CHF) problem in its application to high power reactor. If CHF occurred, boiling regime changes from effective nucleate boiling to ineffective film boiling, so temperature of the RPV goes up and finally the RPV fails. To solve the CHF problem, here have been a lot of works for IVR-ERVC. In the point of in-vessel heat transfer, Theofanous suggested risk oriented accident analysis methodology which is a combination of probabilistic and deterministic approach. A lot of experiments have been done using simulants of corium in various experimental apparatus. Their simulants were usually water due to simulate large Rayleigh number and natural circulation of corium. IVR-ERVC concept has been researched for a long time. For in-vessel heat transfer, simulants or real corium was used to get a heat flux distribution to the outer wall. And based on those results, ex-vessel cooling has been researched in various geometry to get cooling limit as CHF. Material flooding is suggested as improvement of ERVC in APR 1400 to secure safety margin for CHF. Regardless of Prandtl number of the flooding material, the focusing effect of heat flux was mitigated; the maximum heat flux was reduced less than half of the maximum heat flux in bare condition

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.)

Method of estimating the reactor power distribution

International Nuclear Information System (INIS)

Mitsuta, Toru; Fukuzaki, Takaharu; Doi, Kazuyori; Kiguchi, Takashi.

1984-01-01

Purpose: To improve the calculation accuracy for the power distribution thereby improve the reliability of power distribution monitor. Constitution: In detector containing strings disposed within a reactor core, movable type neutron flux monitors are provided in addition to position fixed type neutron monitors conventionally disposed so far. Upon periodical monitoring, a power distribution X1 is calculated from a physical reactor core model. Then, a higher power position X2 is detected by position detectors and value X2 is sent to a neutron flux monitor driving device to displace the movable type monitors to a higher power position in each of the strings. After displacement, the value X1 is amended by an amending device using measured values from the movable type and fixed type monitors and the amended value is sent to a reactor core monitor device. Upon failure of the fixed type monitors, the position is sent to the monitor driving device and the movable monitors are displaced to that position for measurement. (Sekiya, K.)

Electric power from near-term fusion reactors

International Nuclear Information System (INIS)

Longhurst, G.R.; Deis, G.A.; Miller, L.G.

1981-01-01

This paper examines requirements and possbilities of electric power production on near-term fusion reactors using low temperature cycle technology similar to that used in some geothermal power systems. Requirements include the need for a working fluid with suitable thermodynamics properties and which is free of oxygen and hydrogen to facilitate tritium management. Thermal storage will also be required due to the short system thermal time constants on near-time reactors. It is possbile to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system

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)

The future of nuclear power and fourth-generation reactors

International Nuclear Information System (INIS)

Carre, F.; Renault, C.

2006-01-01

Faced with the exhaustion of fossil fuel resources, the output of existing nuclear power must quadruple between now and 2050, and the Commissariat a l'Energie atomique (CEA) and its industrial partners are cooperating in a programme of R and D on future nuclear power. France strategy puts rapid neutron reactors (RNR) at the forefront, in view of their possible introduction by 2040. These reactors allow a more efficient use of uranium resources and minimise the production of long-life nuclear waste. Two technologies which use respectively, sodium and gas as their coolant are being studied. For the sodium RNR, which benefits from significant existing experience, the key is to first improve its economic performance. For the gas RNR, which draws on the principles and the generic assets of the RNR, for those using helium as the coolant, and those with applications at high temperature, it is important firstly to demonstrate the key technologies such as the fuel. The decision of President Chirac to launch the study of a prototype, fourth-generation reactor for 2020 is stimulating the research effort into France future nuclear power. (author)

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

Science.gov (United States)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

SPV Analysis of CEDMCS in Advanced Power Reactors

Energy Technology Data Exchange (ETDEWEB)

Awwal, Arigi M.; Emmanuel, Efenji A. Emmanuel; Faragalla, Mohamed M.; Lee, Yong-kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

Single Point Vulnerability (SPV) is a component whose failure would directly cause an automatic or manual reactor scram or turbine trip. Although some power plants do not consider the cause of any reduction in power as SPV, others consider components that cause a reduction in power of as low as 2% as SPV. The Control Element Drive Mechanism Control System (CEDMCS) controls and regulates power supplied to drive the control rods with the Control Element Drive Mechanism (CEDM). A 4-coil CEDM is used in the newly built Advanced Power Reactor (APR) 1400 plant, while a new CEDMCS for 3-coil CEDM has been designed to be deployed to another APR1400 plant. This paper shows an approach to evaluate the SPVs that may be available in either of these two systems. System A design has employed a fail-safe concept to its design with less redundancies while System B design provides redundancy and design change although this comes at a high price for the Utility. The System B design has improved reliability but not necessarily eliminating the SPV items. Naturally, the cost of a new redundant system will be more. However, future work will examine the economic effect of the new system considering the operating experiences of power plants on the CEDMCS (i.e. SCRAM rates and power outage cost)

The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

International Nuclear Information System (INIS)

Junkrans, S.; Vesterlund, G.; Vaernild, O.

1980-01-01

The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

Transients in reactors for power systems compensation

Science.gov (United States)

Abdul Hamid, Haziah

This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

International Nuclear Information System (INIS)

Server, W.L.; Nanstad, R.K.; Odette, G.R.

2012-01-01

The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

A high temperature reactor could be used to eliminate the Russian military plutonium

International Nuclear Information System (INIS)

Foucher, N.

1999-01-01

The GT-MHR reactor (Gas Turbine Modular Helium Reactor) aims the double objective to eliminate the Russian plutonium coming from weapons, ( until 3 tons by year) and to produce a competitive energy from a small-scale power reactor with a nuclear fuel that can be of different type (plutonium or uranium). This reactor has several advantages: a high yield (47%) as every high temperature reactor and to be used in combined cycle, a high level of safety because of its ability to evacuate the residual power in a totally passive way and because of the nature of its fuel that is made of ceramics with a very high melting point that is to say no possibility of core melt. The fission products are contained in the ceramics so that reactor cannot disseminate radioactivity in its structure and consequently does not induce irradiation for the personnel. (N.C.)